BECO-LTR-95-126, Forwards Info Re Facility Requested by German Federal Ofc for Radiation Protection.W/11 Oversized Drawings

Forwards Info Re Facility Requested by German Federal Ofc for Radiation Protection.W/11 Oversized Drawings

ML20095K444
Person / Time
Site:	Pilgrim
Issue date:	12/19/1995
From:	Olivier L BOSTON EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20095K447	List: BECO-LTR-95-126, Forwards Info Re Facility Requested by German Federal Ofc for Radiation Protection.W/11 Oversized Drawings ML20095K456 ML20095K464
References
BECO-LTR-95-126, NUDOCS 9512290019
Download: ML20095K444 (197)

Text

.. -.

Boston Edison Pilgr.m Nuclear Power Station

" Rocky Hill Acad Pivmouth, Massachusetts 02360 December 19, 1995 5 L J. Olivier BECo Ltr. #95- 126

• Vice President Nuclear Operatons and Station Director U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 1 Docket No. 50-203 L

License No. DPR-35

{_

information Reauested by German Federal Office for Radiation Protection Please find enclosed the information relating to Pilgrim Nuclear Power Station requested by the German Federal Office for Radiation Protection. The request pertained to plant design relating to

<he ventilation system and emission data from 1973 to 1979. The plant design data is comprised of applicable sections of the plant Final Safety and Analysis Report and plant ventilation drawings. The emission 4'ta is comprised of applicable pages of the semi-annual reports map 2

to the NRC during the p of concern. Additionally enclosed are copies of followup epidemiology studies th .y be of interest to the German authorities. An index to the

=

encicsures tollows.

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Please contact Jeffrey Calfa at (508) 830-8108 for questions relating to the plant design data and this letter in general and Thomas Sowdon at (508) 830-8834 for questions relating to the emissions data and the epidamiology followup studies. Thank you.

L W 9512290019 9"1219 p

DR ADOCK 0500 3 I( '

V L.J. Olivier JPC/dmc/ german g cc: Mr. R. Eaton, Project Manager Division of Reactor Projects - l/il Mail Stop: 14D1 U.S. Nuclear Regulatory Commission 1 White Flint North i 11555 Rockville Pike Rockville, MD 20852 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road n~, ,, y King of Prussia, PA 19406

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_ ff Senior Resident inspector Pilgrim Nuclear Power Station J s C._ 'dllA $ hf8O A) bl/kiL t /U

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DCC Desk l December 19, 1995 Page 2 . j Index ofInformation Provided Ikm Descriotion FSAR Section 5.1 Containment Summary FSAR Section 5.2 Primary Containment System FSAR Section 5.3 Secondary Containment System FSAR Section 9.4 Gaseous Radwaste System FSAR Section 10.9 Heating, Ventilation, and Air Conditioning Drawing M210 Main Condenser Air Ejector and Offgas System Drawing M254 Augmented Offgas System Drawing M283 Secondary Containment isolation Control Drawing Drawing M287 Plant Ventilation Drawing Drawing M288 Turbine Building Air Flow Diagram Drawing M289 Reactor Building Air Flow Diagram Drawing M290 Radwaste Area Air Flow Diagram Drawing M294 Standby Gas Treatment System Control Diagram Pilgrim Semi-Annual Radioactive Efiluent Applicable pages demonstrate gaseous Release Reports (applicable pages) from efIluent releases from the plant during the July 1972 to June 1979 applicable time frame Review of the Southeastern Massachusetts Follow-up epidemiology study llealth Study, dated October 1992 Plausib:lity of the Results from the Follow-up epidemiology study Southeastern Massachusetts Health Study 2

9 L PNPS-FSAR pFM r

SECTION 5

. .. A j CONTAINMENT i Q'g'ggj 10 I 5.1

SUMMARY

DESCRIPTION i

5.1.1 General l

The containment systems of Pilgrim Nuclear Power Station utilize a "m.ntibarrier" concept which consists of two systems. The Primary Containment System (PCS) is a pressure suppression system which forms the first barrier. The Secondary containment System (SCS) is- a system which minimizes the ground level release of airborne radioactive materials, and forms the second barrier The fuel, fuel cladding, and Reactor Primary System (RPS) form additional barriers to the release of fission products and are described in Section 3.2.

5.1.2 Primary Containment System The PCS houses the reactor vessel, the Reactor Coolant Recirculation System and other branch connections of the Reactor Coolant System (RCS). The primary containment is a pressure suppression system consisting of a drywell, pressure suppression chamber which stores a large volume of water, a connecting vent system between the drywell and water pool, isolation valves, vacuum relief system, containment

( cooling systems, and other service equipment. The drywell is a steel pressure vessel in the shape of a light bulb, and the pressure suppression chamber is a torus shaped steel pressure vessel located below and encircling the drywell.

The PCS is designed to withstand the forces from any size Et ach of the nuclear system primary barrier up to and including an instantaneous circumferential break of the reactor recirculation piping, and provides a holdup time for decay of any radioactive material released. The PCS also stores sufficient water to condense the steam released as a' result of a breach in the nuclear system primary barrier and to supply the Core standby Cooling Systems (CSCS).

5.1.3 Secondary Containment System The SCS encloses the PCS, the refueling and reactor servicing areas,

~

l new and spent fuel storage facilities, and other reactor auxiliary systems. The SCS serves as the only containment during reactor refueling and maintenance operations, when the primary containment is open, and as an additional barrier when the PCS is functional. The SCS consists of the reactor building, Standby Gas Treatment System 1 (SGTS), main stack, Reactor Building Isolation and Control System (RBICS), and other service equipment.

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5.1-1

PNPS-FSAR The SCS is designed to withstand the maximum postulated seismic )

event, and be capable of providing holdup treatment, and an elevated release -point for any fission products released te it. In addition, the Reactor Building is designed to provide protection for the engineered safeguards and nuclear safety systems located.in the building from all . postulated environmental events including tornadoes. .

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o 5.1-2

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INFORMATION

'h I i 5.2 PRIMARY CONTAINMENT SYSTEM Use restricted to 5.2.1 Safety Objective _-. referenCO __ ,

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- The safety objective of the Primary Containment System (PCS) is to provide the capability in conjunction with other safeguard features, to limit the release of fission products in the event of a postulated

' design basis accident (DBA) so that offsite doses would not exceed 1 the guideline values set forth in 10CFR100.

$ 5.2.2 Safety Design Basis

! 1. -The PCS shall have the capability of withstanding the conditions which could result from any of the postulated DBAs for which the PCS is assumed to be functional, including the largest amount of energy release and mass flow associated with the accident.

2. 'The PCS shall have a margin for metal water reactions and other chemical reactions subsequent to any postulated DBA for which the PCS .is assumed to be functional, consistent with the performance objectives of the nuclear safety systems and engineered safeguards.

3. The PCS shall have the capability to maintain its functional integrity during any postulated external or environmental event.

4. The PCS shall have the capability to be filled with water as an accident recovery method for any postulated DBA in which a breach of the nuclear system primary barrier cannot be sealed.

5. The PCS, in conjunction with other Nuclear Safety Systems and engineered safeguards, shall have the capability to limit leakage during any of the postulated DBAs for which it is assumed to be functional, such that offsite doses do not exceed the guideline values set forth in 10CFR100.

6. The PCS shall have the capability to rapidly isolate all pipes or ducts necessary to establish the primary containment barrier.

7. The PCS shall have the capability to store sufficient water to supply the Core Standby Cooling System (CSCS) requirements.

8. The primary containment shall have the capability to be maintained during normal operation within the range of initial conditions assumed in the Station Safety Analysis in Section 14.

5.2.3 Description 5.2.3.1 General The design employs a Low Leakage Pressure Suppression Containment System which houses the reactor vessel, the reactor coolant recirculating loops, and other branch connections of the Reactor Primary System. The Pressure Suppression. System consists of a 5.2-1

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

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PNPS-FSAR  :]

l a pressure suppression _ chamber (torus) which stores a large j idrywell,

1. l volume:of water, a' connecting vent system.between the drywell_and the 1 l -jpressure suppression pool, isolation valves, Vacuum, Relief System,-

' Containment' Cooling Systems, and other. service equipment. The PCS design' parameters are given on Table 5.2-1.

) In the event of a Process System piping failure within the drywell,'

reactor water and steam will be released into the drywell gas space.

The 'resulting increased drywell pressure forces a mixture of air, ,

steam,- and ' water 'through the vent system ~into the pressure  ;

suppression pool. The steam condenses rapidly in the suppression l

. pool resulting in rapid pressure reduction in the drywell. Air transferred during reactor blowdown to the suppression chamber

- pressurizes the chamber, and subsequently is vented to the drywell '.

through the vacuum relief system as the pressure in the drywell drops below that in the suppression chamber.

Cooling systems are provided to remove heat from the water in the ,

suppression chamber. This provides for continuous cooling of the i primary containment under the postulated DBA conditions for which the PCS is assumed to be functional. Isolation valves are provided to ensure containment of radioactive materials within the primary .

containment, which might be released from the reactor to the l during the course of an accident. Other service  !

containment equipment is provided to maintain the containment within its design l parameters during normal operation.

The drywell (primary containment) coolers are designed to maintain i drywell atmosphere temperatures within an acceptable range during normal station operation. See Table 5.2-2. The reduction of atmosphere temperature by the coolers will also result in partial condensation of water vapor when the incoming humidity levels are high.

The drywell fan motors are rated for continuous operation in  :

atmospheres having 100 percent rh and 104*F temperatures. In the  !

design of the cooler, the motor has been placed in the exhaust of the cooler where the leaving air temperature is a maximum of 95 F, so ,

that the motor is exposed to the lowest humidity and lowest temperature atmosphere available within the drywell. Pressure increases to the 2.5 psig high drywell pressure condition used to sense a possible loss of coolant accident (LOCA) would not affect the l continued operability of the coolers. The drywell coolers are automatically shut down in the event of a LOCA combined with the loss of offsite ac power.

The drywell coolers, including the fans, with their power and control systems were tested during the preoperational tests at the station to  !

demonstrate the required operability of the power and control i systems, the fans, and'the Reactor Building closed cooling water l supply.to the coolers. The capability of the coolers to maintain the required drywell atmosphere temperatures was verified during the startup program as the drywell heat loads increased during the heatup and pressurization of the Nuclear Steam Supply System.

1 5.2-2

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PNPS-FSAR The Reactor. Building Closed Cooling Water System (RBCCHS) piping-supplying the drywell coolers will be revised to seismic Class I to maintain the pressure boundary integrity of this piping under seismic loading. Refer to _ Section 10.5.5.1. The drywell coolers were originally purchased as- seismic Class I equipment to serve-as pressure boundary only.

The PCS design loading considerations are given .in Section 12 and Appendix C. The Station Safety Analysis presented in Section 14 demonstrates the effectiveness of the PCS as a radiological barrier.

In adcition, primary containment pressure and temperature transients from ustulated DBAs are also presented in Section 14.

5.2.3.2 Drywell The drywell is a steel pressure vessel with a spherical lower portion, 64 ft in diameter, and a cylindrical upper portion 34 ft 2 inches in diameter. The overall height is approximately 110 ft. The design, fabrication, inspection, and tesung of the drywell vessel complies with requirements of the ASME Boiler & Pressure Vessel Code,'Section III, Subsection B, Requirements for Class B Vessels, which pertain to containment vessels for nuclear power stations.

The drywell structure is designed for an internal pressure of 56 psig l coincident with a temperature of 281*F with applicable dead, live, and seismic loads imposed on the shell. Thus, in accordance with the ASME Code,Section III, Code Case N-1312-(2), the maximum drywell pressure is 62 psig. Thermal stresses in the steel shell due to temperature gradients are taken into account in the design.

Special precautions not required by codes were taken in the fabrication of the steel drywell shell. Charpy V-notch specimens were used for impact testing of plate and forging material to give i assurance of proper material properties. Plates, forgings, and pipe

! associated with the drywell have an initial NOT temperature of 0*F or lower when tested in accordance with the appropriate code for the materials. It is intended that the drywell will not be pressurized or subjected to substantial stress at temperatures below 30*F. ,

! The drywell is enclosed in reinforced concrete for shielding purposes,

! and to ' provide additional resistance to deformation and buckling in

! areas wherc the concrete backs up the steel shell. Above the '

' transition zone, the drywell is sep,arated from the reinforced concrete by a gap of approximately 2 in. Shielding over the top of the drywell is provided by removable, segmented, reinforced concrete shield plugs.

In addition to the drywell head, one double door air lock and two .

' bolted equipment hatches are provided for access to the drywell. The

' locking mechanisms on each air lock door are designed so that a tight

! -seal ~will be maintained when the doors are subjected to design pressures. The doors are mechanically interlocked so that neither

door may be operated unless the other door is closed and locked. The

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p 5.2-3 Revision 14 - June 1992  ;

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PNPS-FSAR drywell head and equipment ha'tch covers are bolted in place and sealed with gaskets.

The spectrum of primary system leak . rates up to a double ended blowdown of a' recirculation line has been analyzed relative to the Steam issuing from temperature and pressure response of the' drywell.

a leak and expanding.at constant enthalpy may result in a superheated ,

containment atmosphere. The maximum amount of superheat possible is a function of both the source pressure (reactor pressure) and~ the '

receiver pressure (drywell). ~ The enthalpy of saturated steam. goes '

' through a maximum value at a reactor pressure. of 400 to 500 psia.

Steam issuing from a leak at this pressure will result in the maximum superheat for a given containment pressure.

If a -steam leak occurs, the containment pressure and temperature increase at a rate dependent .on the size of the leak, containment characteristics, and the pressure of the reactor. The containment ,

pressure and temperature rises as noncondensable gases are swept into the suppression chamber. Containment pressure levels off after The all noncondensable gases are driven into the suppression chamber.

on the containment shell temperature rises as steam condenses relatively cool wall. When the drywell shell temperature reaches the saturation temperature dictated by this containment pressure, steam condensation is terminated. The o'dy energy available to further increase the wall temperature is the superheat energy. The result is

'a decrease in the rate of temperature rise of the drywell shell and an increase in the bulk atmosphere temperature of the drywell.

Figure 5.2-1 illustrates the reactor vessel Dressure response to steam leaks ranging in size from 0.02 to 0.50 ftj Figures 5.2-2 through 5.2-6 illustrate the containment response to steam leaks covering the same size range. The response of the containment to small steam leaks i

is slow, but the continued high reactor pressure results in high containment temperature, given enough time. Leaks so small that the high drywell pressure trip does not occur will not result in a high 7

l temperature. Leaks large enough 'to result in a high containment j temperature will be large enough to sweep air into the suppression chamber and result in significant drywell pressure increase. Large

! leaks will either depressurize the reactor rapidly or result in auto-relief such that steam temperatures, reaching levels up to 330*F, will not persist long enough to result in structural wall temperatures exceeding 281*F.

Safety grade temperature monitoring instrumentation is provided ir the Control Room so that tctivation of one of the two containment sprays would be effective it' terminating the temperature rise because the superheat is quickly removed. The spray nozzles are designed to give a small particle size, and the heat transfer to the subcooled spray is very effective. Sinte the total amount of heat in the drywell j atmosphere is low ralative to the spray rate, the containment

[ atmosphere temperature is quickly reduced to near the spray temperature.

5.2-4 Revision 14 - June 1992 i

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PNPS-FSAR A drywell pressure condition exceeding 10 psig was selected as the

' basis for determining when to initiate the containment spray. See Figure 5.2-7 for time required to reach 10 psig. The operator will be instructed to initiate the containment sprays if containment ' pressure exceeds 10 psig for longer than 30 min. Safety grade pressure

monitoring instrumentation is provided in the Control Room for this purpose. This procedure will ensure that the containment wall never exceeds 281*F. Depressurization of the reactor vessel can take place at - the normal rate, but depressurization is not required to ensure that the wall temperature remains below 281*F. The environmental conditions considered in the design of the reactor protective system instrumentation, engineered safety feature equipment, and the

qualification tests that have been conducted are described in Section i 7.1.8. The analyses of steam leaks inside the containment is given in

- detail in references 7 and 8.

5.2.3.3 Pressure Suppression Chamber and Vent System i

i l 5.2.3.5.1 General The pressure suppression pool, which is contained in the pressure

suppression chamber, initially serves as the heat sink for any postulated transient or accident condition in which the normal heat sink, main condenser, or Shutdown Cooling System is unavailable.

Energy is transferred to the pressure suppression pool by either the

discharge piping from the reactor pressure relief valves or the j Drywell Vent System. The relief valve discharge piping is used as the d- energy transfer path for any condition which requires the operation of 1 the relief valves. The Drywell Vent System is the energy transfer ,

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. path for all energy releases to the drywell.

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Of all the postulated transient and accident conditions, the

instantaneous circumferential rupture of the reactor coolant 4 iecirculation piping represents the most rapid energy addition to the

!~ pool. For this accident the vent system, which connects the drywell and suppression chamber, conducts flow from the drywell

~

to the j suppression chamber without excessive resistance and distributes this 1 flow effectively and uniformly in the pool. The pressure suppression pool receives this flow, condenses the steam portion of this flow, and releases the noncondensable gases and any fission products to the 4

pressure suppression chamber air space.

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5.2.3.3.2 Pressure Suppression Chamber

). The pressure suppression chamber is a steel pressure vessel in the

shape of a torus below and encircling the drywell, with a centerline j

vertical pressuredia of 29 ft 6 chamber suppression in and a contains horizontal dia of 131 ft84,000 approximately 6 in. 3ft The of I

water and pasThe 120,000 ft .

a net air space above the water pool of approximately suppression chamber will transmit seismic loading to the reinforced concrete foundation slab of the Reactor Building. Space l

is provided outside of the chamber for inspection. j

! 5.2-5 Revision 14 - June 1992 4

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The toroidal suppression chamber-is designed to the same material- and ,

code - requirements as the steel ' drywell - vessel. The material has an i ~

o NDT temperature of_O'F or less.

1 5.2.3.3.3 -Vent System-j Large vent pipes- connect the drywell and the pressure suppression j chamber. A total of eight circular vent pipes are provided, each 4 having a. dia of 6.75 ft.. The vent pipes are designed for the same >

i- pressure - and temperature conditions as the drywell and suppression chamber.. Jet deflectors are provided in the drywell at the entrance of each vent - pipe _. to prevent possible damage to the vent pipes from

[~ The jet forces which might accompany a pipe break in the drywell. with vent . pipes are fabricated of SA-516 steel, and comply requirements of the ASME Boiler and Pressure Vessel Code,.Section III, l

i - Subsection B. The vent pipes are provided with expansion joints which motion

• are enclosed within sleeves, to accommodate differential l' - between the drywell and suppression chamber. l i

i- The drywell vents are connected to a 4 ft 9 in dia vent header in the t l

form of a torus which is contained within the airspace of the L suppression chamber. Projecting downward from the header are 96 ,

i downcomer pipes, 24. inches in dia, terminating approximately 3.00 to t 3.25 ft below the water surface of the pool. The vent header has the ,

same temperature and pressure design requirements as the vent pipes.

t Vent pipes and vent headers are braced to withstand expected loads from steam blowdown into the pool.

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'5.2.3.3.4 Pressure Suppression Pool l The pressure suppression pool is approximately 84,000 ft3 of demineralized water contained within the pressure suppression chamber. It serves both as a heat sink for postulated transients and +

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. accidents and as a tource of water for the CSCS. i l The suppression pool receives energy in the form of steam and water from the reactor pressure relief valve discharge piping, or the i

drywell vent system downcomers which discharge under water. The steam

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is condensed by the suppression pool. The condensed steam and any l

water carryover cause an increase in pool volume and temperature.

Energy can be removed from the suppression pool when the Residual Heat Removal System (RHRS) is operating in the suppression pool cooling mode.

The suppression pool is the primary source of water for the Core Spray and Low Pressure Coolant Injection (LPCI) Systems, and the secondary ,

i source of water for the Reactor Core Isolation Cooling (RCIC) and High Pressure Coolant Injection (HPCI) Systems. The water l

5.2-6 Revision 10 - July 1989 l l

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i PNPS-FSAR level and temperature of .the : suppression _ pool are continuously monitored in the main control room.

5.2.3.4 Penetrations

-5.2.3.4.1 General Containment penetrations have the following design characteristics:

1. They are . designed for the same pressure and temperature-conditions as the drywell and pressure suppression chamber

2. They are capable of withstanding the forces caused by impingement of the fluid from the rupture of the largest local pipe or connection without failure

3. They are capable of accommodating the thermal and mechanical stresses, which may be encountered during all modes of operation including environmental events, without failure

4. They are capable of withstanding the maximum reaction that the pipe to which they are attached is capable of exerting The penetration schedule, including the number and size of these penetrations, is shown on Table 5.2-3. Load combinations and allowable stresses are described in Appendix C.

5.2.3.4.2 Pipe Penetrations Two general types of pipe penetrations are provided. Type 1 is useo where the design must accommodate thermal movement. Figure 5.2-9 is typical of this type of penetration. Type 2 is used where stresses due to thermal movement are relatively small. Typical penetrations of this type are illustrated on Figures 5.2-10 and 5.2-11.

Figure 5.2-12 shows a typical instrument penetration.

The piping penetrations which have special provisions for thermal movement, such as the steam lines, are shown on Figure 5.2-9. In these penetrations, the process line is enclosed in a guard pipe that is attached to the main steam line through a multiple head fitting.

This fitting is a one-piece forging with integral flues or nozzles and is designed to meet all requirements of the ASME Boiler and Pressure Vessel Code, Section III, Class B. The forging is radiographed and ultrasonically tested as specified by this code.l The guard pipe and flued head are designed to the same pressure requirements as the process line. The process line penetration sleeve is welded to the drywell, and extends through the biological shield where it is welded to a two-ply expansion bellows assembly, >

which in turn is welded to the flued head fitting. The pipe is guided through pipe supports at the end of the penetration assembly to allow steam pipe movement parallel to the penetration, and to limit pipe reactions of the penetration to allowable stress levels. ,

5.2-7 Revision 2 - July 1983

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.FHPS-FSAR r

Where -necessary, the penetration assemblies are anchored outside the-

-containment to limit' the movement of the line relative to the

. containment. The bellows accommodates the relative movement between the pipe and the containment shell.

The -design of' the penetration takes into account the stresses associated with normal thermal expansion, live and dead loads, seismic , loads, and loads associated with a LOCA within the drywell.

The design takes into account.the ltadings given above in addition to the'. jet force loadings resulting from any pipe _ failure. .The resultant stresses;in the pipe and penetraticn for the condition do not exceed 90 percent of the material yield stress.

The cold piping, ventilation duct, and instrument line penetrations-

-are generally welded directly to the sleeves. In some cases, where

. stress analyses indicate the need, double flued head fittings are used. Bellows and guard pipes are not necessary in these designs, since the thermal stresses are small and are accounted for in the

' design of the weld joint.

5.2.3.4.3 Electrical Penetrations The electrical penetrations include electrical power, signal, and instrument leads. Typical electrical penetrations are shown on Figures 5.2-13, 5.2-14, and 5.2-15. The penetrating sleeve is welded to the primary containment vessel. Medium voltage power penetrations ,

primary seals are made of alumina-ceramic materials. The seals are formed at 1,300 F or higher, and thus the temperatures to which the seals would be exposed during a LOCA would have no adverse effect on their leaktightness characteristics.

The electrical penetrations used for low voltage power, control, and instrumentation cable and for coaxial cable utilize either A1202 or a bonding resin to maintain the leaktight integrity of the containment penetrating sleeves. A prototype of the penetration assembly which utilizes a bonding resin has been tested by exposing the interior face of the penetration assembly to the following environmental conditions: 281*F, 63 psig internal pressure,90-100 percent rh for 10 days. An additional test at 320 F, 125 psig internal pressure and 90-100 percent rh for 2 hr was conducted. The pressure retaining t capability of the penetration assembly was maintained throughout the duration of the tests.

'- The leak rate was monitored during the test and did not exceed 24 cc/hr of nitrogen through the inner seal. The outer seal is not exposed to high temperatures during an accident and therefore the ,

overall leak rate through both seals is 10-6 cc/sec. ,

> Additional tests were planned to certify the pressure retaining capability of those penetrations utilizing bonding resin at 340 F, 100 percent rh for 30 min. j A prototype of the penetrations using a polysulfone seal has been qualified to the following environmental conditions: 340 F, 110 psig i 5.2-8 i

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PNPS-FSAR for 6 hr; 320*F, 75 psig for 3 hr; and 260*F, 20 psig for 12 days. The inboard 5.3 x 10gnd/or outboard seal possessed a leak rate that was less than cc/sec helium.

Section 7.1.8 states that qualification tests were to be conducted on the medium voltage electrical penetrations, including leakage tests, following environmental exposures in excess of the design basis LOCA conditions.

5.2.3.4.4 Traversing Incore Probe Penetrations Traversing incore probe (TIP) guide tubes pass from the Reactor Building through the primary- containment. Penetration of the guide tubes through the primary containment are sealed by means of brazing which meets the requirements of the ASME Boiler and Pressure Vessel Cede,Section VIII. These seals would also meet the intent of Section III of the code even though the code has no provisions for qualifying the procedures or performances.

5.2.3.4.5 Personnel and Equipment Access Locks One personnel access lock is provided for access to the drywell. The lock has two gasketed doors in series, and each door is designed to withstand the drywell design pressure. The doors are meenanically interlocked to ensure that at. least one door is locked at all times when primary containment is required. The locking mechanisms are designed so that a tight seal will be maintained when the doors are subjected to either the design internal or external pressure. The seals on this access opening are capable of being tested for leakage.

A personnel access hatch with testable seals is provided on the drywell head. This hatch is bolted in place.

l Two equipment access hatches with testable seals are also provided. l l These hatches are bolted in place.

5.2.3.4.6 Access to the' Pressure Suppression Chamber Access to the pressure suppression chamber is provided at two locations from the Reactor Building. There are two 4 ft dia manhole entrances with double gasketed bolted covers connected to the chamber by 4 ft dia steel pipes.

5.2.3.4.7 Access for Refueling Operations The drywell vessel head is removed during refueling operations. The I head is held in place by bolts and is sealed with a double-seal  !

arrangement. )

5.2-9

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5.2.3.5: Primary Containment Isolation Valves j i

5.2.3.5.1 _ General Criteria The: basic function of all ' primary containment isolation valves is to provide necessary isolation to the containment -in. the event' of accidents or similar critical conditions ' when The the free release:

containment of isolation containment atmosphere cannot be permitted.

valves are listed on Table 5.2-4. This table also defines the -valve status (normally open or normally closed) during normal reactor operation and shows the signals required to initiate their. desired l operation. The primary containment isolation valves are grouped into four basic classes..

Class A valves are on process lines that communicate directly with the reactor vessel and penetrate the . primary containment. These lines require ~two valves in series, one inside the primary containment and one outside the primary containment. They are located as close to the primary containment boundary as practical. Except in the case of check valves, both valves shall close automatically on isolation _ signal.

Both valves shall ' receive the isolation (closure) signal even if normally closed during reactor operation. Since deck valves close on reverse process flow, they are used to isolate some incoming lines. ,

All Class A valves except check valves are capable of remote manual control from~the control.toom.

Class B valves are on . process . lines that do not directly communicate '

with the reactor vessel, but penetrate the primary containment and communicate with the primary ' containment free space. These lines require two valves, in series, both of them located outside the primary containment, and as close to the primary containment boundary as practical. Except in the case of check valves, both valves close automatically on isolation signal. Both valves receive the isolation closure signal even if normally closed during reactor operation. See Table 5.2-4 for valve status during reactor operation. All Class B

- valves except check valves are capable of remote manual control from the control room.

Class C valves are on process lines that penetrate the primary containment but do not communicate directly with the reactor vessel, with the primary containment free space, or with the environs. Class C lines require only one valve which closes automatically by process action (i .e. , reverse flow) or by remote manual operation from the control room (Reference 6, Section 5.2.9). l Motive power for the valves on process lines which require two valves shall be from physically . independent sources to provide a high probability that no single accidental event could interrupt motive power'to,both closure devices.

5.2-10 ' Revision 13 - June 1991

PNPS-FSAR Variations .to the above definitions are referenced on Table 5.2-4 by

' their class designations followed by an "X" suffix.- The lines in this class are generally instrument lines or lines used for. core cooling.

. Automatic isolation valves, in the usual sense, are not used 'on the inlet lines of the Reactor Core and Containment Cooling ' Systems and Reactor Feedwater Systems, since operation of these systems is essential following a design basis LOCA. Since normal flow of water in these systems is inward to the reactor . vessel or to the primary containment, check valves located in these lines will provide automatic isolation, if necessary.

No automatic isolation valves are provided on the Control Rod Drive System hydraulic lines. These lines are isolated by the normally closed hydraulic system control valves located in the Reactor Building, and by check valves comprising a part of the drive mechanisms.

TIP lines and small diameter instrument lines are not provided with automatic isolation valves.

5.2.3.5.2 Additional Considerations Effluent lines such as main steam lines, which connect to the reactor vessel or which are open to the primary containment, have air-powered valves. This arrar.gement provides a high reliability with respect to functional performance. These valves are closed automatically by the signals indicated on Table 5.2-4.

The HSIV's are also connected to the nitrogen supply system. This 3 redundant source of MSIV actuation results in greater system reliability.

TIP system guide tubes are provided with an isolation valve which closes automatically upon receipt of proper signal and after the TIP cable and fission chamber have been retracted. In series with this isolation valve, an additional or backup isolation shear valve is included. Both valves are located outside the drywell. The function of the shear valve is to assure integrity of the containment in the unlikely event that the other isolation valve should fail to close or the chamber drive cable should fail to retract if it should be i

4 extended in the guide tube during the time that containment isolation is required. This valve is designed to shear the cable and seal the  ;

guide tube upon an actuation signal. Valve position (full open or full closed) of the automatic closing valves will be indicated in the control room. Each shear valve will be operated independently. The valve is an explosive type valve and each actuating circuit is monitored. In the event of a containment isolation signal, the TIP system receives a command to retract the traveling probes. Upon full

[ retraction, the isolation valves are then closed automatically. If a traveling probe were jammed in the tube run such that it could not be retracted, instruments would supply this information to the operator, who would in turn investigate to determine if the shear valve should be operated.

5.2-11 Revision 10 - July 1989

PNPS-FSAR The t'wo 18 in purge and vent line pipe entrances into the drywell have  ;

been provided with baffle plates to prevent debris from entering the lines during an accident. Any debris would threaten the ability to close the applicable isolation valve. 5 The N2 makeup 4nd vent isolation valves are used to relieve' high However, these valves drywell pressure during nonaccident conditions.

may be used after an accident provided the required power supplies are Section available and a low-low water . level signal is not present. -

- 5.4.3 - describes the N2 makeup and vent valves used following an accident condition. ,

Lines .such as those of the RBCCHS which do not connect to the Reactor  !

Primary System or open into the primary containment, are provided with at least one ac-powered valve on the effluent line and a check valve on the influent line.

The Control Rod Hydraulic System is provided with three valves which can be utilized for isolation purposes. The first is a ball check.

valve which comprises an internal portion of the control drive mechanism. The other valves are normally closed hydraulic system control valves located in the Reactor Building.

5.2.3.5.3 Instrument Piping Connected to the Reactor Primary System Instrumentation piping connecting to the Reactor Primary System which L leaves the primary containment is dead-ended at instruments located in the Reactor Building. These lines are provided with flow limiting i

orifices, manual isolation valves, and excess flow check valves, j Instrument sensing lines that originate within the reactor coolant ,

1 pressure boundary and penetrate the primary containment are 1 in dia '

seismic Class I lines; 1/4 in dia orifices are installed in each of these lines inside the primary containment. This orifice size was i selected to provide the same effective fluid cross sectional area as

the excess flowcheck valves when fully open. A manually operated stop

! valve and excess flowcheck valve are installed in each line 4

immediately outside, and as close as practicable to the primary containment consistent with the requirement for access to the stop j_ valve. The combination of orifice and excess flowcheck valve will l

reduce leakage to as low a value as practicable in the unlikely event

of line failure. A line failure downstream of the excess flow check valve will result in a maximum leakage rate of 2 gpm prior to l ;

actuation. A failure of the excess flowcheck valve body or the j' instrument line upstream of this valve would result in a maximum

. leakage rate of 20 gal / min. In each of these instances the leakage is '

well within the capability of the Reactor Coolant Makeup System.

4 f.

5.2-12 Revision 14 - June 1992

PNPS-FSAR

" The- amnu.nt of steam- released to the- Reactor Building from' a : i

20. gal / min leak would not result in a failure' of- secondary containment.. If the Reactor - Building is not' .' isolated, there would l not be any significant -pressure rise due to the relatively high

' Reactor Building ventilation ' exhaust rates. If the Reactor Building is isolated, the operation .of one standby gas . treatment filter train will prevent Reactor Building pressure from' exceeding'.its design value. -[

'An analysis of the potential offsite exposure that would result from a 20 gal / min leak into the Reactor Building has been performed, Such a : leak corresponds to an assumed failure of an instrument line outside the primary containment but upstream of . the excess flowcheck valve. It was assumed . in the analysis ' that _ manual shutdown and depressurization would be- initiated within 30 min. The delay 'of I

l f.

t i

1 L ,

p 5.2-12a Revision 4 - July 1984 i f-i i

PNPS-FSAR 30 min is extremely conservative considering the numerous ways such a leak may be detected.

The analysis assumed that steam from the leak would be released to the environment through the normal ventilation path until the reactor had been depressurized. Based on these assumptions, the total dose at the site boundary for the duration of exposure was computed to be 0.15 rem to the thyroid, which is substantially below the guidelines of 10CFR100.

Pressure retaining welds of instrument sensing lines that are part of the reactor coolant pressure boundary receive magnetic particle or liquid penetrant examination of the last pass.

Instrument line " bundles" are routed so as to minimize the potential for accidental damage. They are generally routed high in compartments to ensure they are not stepped on or otherwise damaged.

The lines are equipped with flow limiting orifices and excess flowcheck valves and are of the same size and schedule; therefore, the possibility of one line causing failure in another is extremely remote.

The containment penetrations for these sensing lines are shown on Figure 5.2-12. The 10 in drywell penetration sleeve contains six, equally spaced, 1 in, schedule 80 stainless steel instrument lines.

The manual isolation valves are 1 in stainless steel globe valves and are located as close to the penetration as practical, consistent with the need for access to the valve. The excess flowcheck valves close automatically on flow in excess of 2 gal / min. Neither the manual isolation valves nor the excess flowcheck valves are equipped with position indicators. Regular monitoring of measured variables and comparison between redundant instruments provides operating personnel with sufficient information to identify malfunctioning or inoperative instruments and sensing lines. Operating and/or testing procedures will assure the operability of the safety related instrument lines and their associated orifices and excess flow check flows.

An analysis was conducted to determine the amount of Reactor Building ventilation that would be required to prevent exceeding the design internal pressure of the Reactor Building for an instrument line blowdown through a 1/4 in orifice. The required ventilation flow rate under these conditions is approximately 2,000 ft 3/ min, which is far below the available flow rate through either the normal Reactor Building Ventilation System or the Standby Gas Treatment System (SGTS). An instrument line failure will therefore not result in a loss of integrity of the Secondary Containment System (SCS).

An estimate of the potential offsite exposure that would result from an instrument line failure has been calculated. The assumptions employed in this analysis were:

1. An instrument line failure occurs and results in an initial blowdown of 2.2 lb mass /sec into the Reactor Building 5.2-13

PNPS-FSAR

2. This blowdown continues undiminished and undetected for a period of 30 min

3. After a period of 30 min, the reactor is shut down, depressurized, and cooled down at a controlled 100'F/hr

4. The water which flashes to steam is carried out of the Reactor Building by the normal ventilation system for the duration of the blowdown

5. The I-131 concentration in the blowdown is 6.1 x 10-2 microcurie /ml and the total iodine concentration is 1.6 x 100 microcurie /ml

6. The atmospheric diffusion factor (X/0) for a ground level release, 500 m distance to site boundary, and wake dilution factor of 3 is 5 x 10-4 sec/m3

7. The breathing rate is 3.47 x 10-4 ma /see The above estimates assume that corrective action would not begin for a period of 30 min. The detection of a sensing line break would 'be almost immediate by one or a combination of the means listed below.

Proper corrective action would then be taken by the operating staff in accordance with station procedures such that the leak would be isolated or station shutdown and depressurization be initiated. It is believed that it is not credible to assume no operator action would be taken in 30 min to terminate the consequences, and that the analysis based on a 30 min allowance for these actions is very conservative.

Sensing line break detection means are:

1. By a scram, annunciation, and possible instrument readouts and/or initiation of reactor safeguards systems if rupture occurred on a Reactor Protection System instrument line

2. By annunciation of the control funccion, either high or low in the control room

3. Operator comparing readings with several instruments monitoring the same process variable such as reactor level, jet pump flow, and steam pressure

4. By increases in area temperature monitor readings and high temperature alarms in the Reactor Building, and/or ventilation exhaust air ducts

5. By a general increase in the area radiation monitor readings throughout the Reactor Building

6. The leak should be audible either inside the Turbine Building or outside the Reactor Building to the operating staff members on a normal tour 5.2-14

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

! PNPS-FSAR

! 7. - By ' detecting the leak as soon as an access door to the Reactor Building is. opened or approached ,

l Routine surveillance and the multiplicity of detection methods on the

" part of the operator as given in items 1 through 7 above, represent an i

adequate means for detection of incipient or sudden failure of these i- small diameter instrument lines and components.

' 5.2.3.6 Venting and Vacuum Relief System ,

4

- 1. General The purpose of the vacuum relief valves is to equalize the ' pressure 4

! between the drywell,and suppression chamber and reactor buildingThe so i

that the structural integrity of the containment is maintained.

vacuum relief system from the pressure suppression chamber to reactor building consists of two 100-percent vacuum relief breakers (2 parallel sets of 2 valves in series). Operation of eitherthe system will maintain the pressure differential less than 2.0 psig: external I ,

design pressure. One valve may be out of service for repairs for a i

period of 7 days. If repairs cannot be completed within 7 days, the reactor coolant system is brought to a condition where vacuum relief  ;

is'no longer required.

I The capacity of the 10 drywell vacuum relief valves are sized to limit

?

the pressure differential between the suppression chamber and drywell during post-accident drywell coolant operations to the design limit of l J l

i 2.0 psig. They are sizef }on the basis of the Bodega Bay pressure  !

suppression system tests The ASME Boiler and Pre sure Vessel )

Code,Section III, Subsection B, for this vessel allow; a 5 psig l

' vacuum; therefore, with two vacuum relief valves secured in the closed position and eight operable valves, containment integrity is not j'  !

i impaired.  !

Reactor operation is permissible if the bypass area between the l primary containment drywell and suppression chamber does not exceed an l

allowable area. The allowable bypass area is based upon analysis j area, suppression chamber considering primary system break effectiveness, and containment design pregsure. Analyses show that i

j the maximum allowable bypass area is 0.2 ft .

Reactor operation is not permitted if the differential pressure decay 4

rate is demonstrated to exceed 25 percent of allowable, thus providing L

a margin of safety for the primary containment in the event of a small 4 break in the primary system.

2. Relief Valve Monitors l- The drywell to torus vacuum breakers are installed to assure that the drywell pressure is at least equal to or greater than the pressure in the torus. In addition, when the vacuum breakers are in thedirected closed 4 position, the drywell atmosphere (postulated steam) is l

. )

-5.2-15 Revision 11 - July 1990 f

_ = _ - _________ - . - . . . ... .. - , .- .

PNPS-FSAR of through the suppression chamber downcomers during conditions drywell pressurization. To. fulfill this engineered safety feature, proper positioning and operation of. the vacuum breakers must be ensured. Therefore, each Pressure Suppression Chamber-Drywell Vacuum Breaker is fitted with redundant pairs of position switches which provide signals of disk position to panel-mounted indicators and redundant annunciators to alarm in the main' control room if the disk is open more than the allowable limit.

5.2.3.7 Primary Containment Cooling and Ventilation System The Primary Containment (drywell) Cooling System utilizes eight fan coil units distributed inside the drywell. See Figure 5.2-18. The Primary Containment Cooling and Ventilation System design parameters are given on Table 5.2-2. Each fan coil unit consists of two coolingEach coils and two direct-connected motor-driven vaneaxial fans.

cooling coil is connected to a cooling water supply and return piping system inside the drywell. One or both cooling coils may be utilized l

l l

5.2-15a Revision 11 - July 1990

PNPS-FSAR

( .

control. . Each unit recirculates the drywell for: ' temperature atmosphere through > the cooling - coils to control the drywell space 4 temperature. Cooling water is supplied from the RBCCWS, l' .Thermocouples are provided to monitor the performance of the drywell j cooling system. .They are installed in the air and water connections

! of the drywell-_ coolers as well as the air outlets of the reactor

,vessel . recirculation pump motors u shown on Figure 5.2-18. (The thermocouples 'on water connections'are also shown on Figure 10.5-1).

}:

i Temperature readouts are provided on indicating panel C-2261A locate i outside the drywell.

L Fan coil units circulate cooled air around the recirculating pumps

- and motors, the control rod drive area, and the annular space between I

the reactor pressure vessel and the biological shield. The personnel access and control rod drive removal openings are sealed to ensure L positive flow of cool air from the control rod drive area into the

. annular space between'the reactor vessel and the' biological shield, j through pipe openings in the reactor vessel support located primarily j at the upper level of the control rod drive spac.e.

Cooled air will also be circulated through the reactor vessel head i area, the space immediately below the refueling seal plate, and the relief valve area.

7 Each fan coil unit-has provisions for installing dust filters. i Filters are to be employed during drywell maintenance activities and l will be removed prior to normal station operation. f Cooling water flow to each coil is controlled independently by an

.- electric motorized modulating valve positioned by a valve positioner in the control room. The cooling coil leaving air temperature can be  ;

e adjusted by regulating the flow of cooling water. A cooling coil l

failure can be detected by a flow device located in the cooling unit i condensate drain line, which is annunciated in the main control room.

j- The standby coil can.be put in service and the other isolated by their motorized valves and a check valve in the return line.

1

}. Each fan is started from a local panel by using run-off-auto type j switches. One fan is started by switching to RUN and the other fan switch is placed in the AUTO position. If the normal operating fan fails, a flow switch will sense a reduced pressure and automatically

> start the standby fan, light an amber light at a local panel, and annunciate in the control room. Cooling unit discharge air

{ temperature is sensed by a temperature element and indicated in the

. control room. Upon scram, standby fans will be placed in service automatically to provide additional cooling. All fan coil units can be operated from the emergency power supplies.

The drywell purge ventilation supply system consists of. two full capacity fans to supply clean Reactor Building air to the drywell for purge and ventilation purposes, during the reactor shutdown and i refueling periods for personnel access and occupancy. The purge exhaust air is normally discharged to the atmosphere through the 4-5.2-16 Revision 4 - July 1984 i

• PNPS-FSAR i

Reactor Building exhaust vent. If necessary, SGTS is used for' cleanup and the drywell air is exhausted through the main stack.

• The ventilation lines supplying the primary containment are provided .

with two fast acting, pneumatic, - cylinder-operated butterfly valves in series for isolation. purposes. These valves are normally closed during station operation.

4 J

i 1

1 5

5.2-16a Revision 4 - July 1984

PNPS-FSAR Procedures. for normal primary containment venting and purging.are established such that gaseous effluent releases from the- station As noted above, purge or remain within ' the normal release limits.

vent exhausts can be directed to elevated release points through. the Reactor Building vent, or through the SGTS to the main stack.

Drywell and torus purging will normally be conducted to facilitate personnel access subsequent to periods of operation with the primary containment inerted.. Primary containment purge operati would standard ftgnsof gas.

normally release on the order of 1 million Drywell and torus venting is required during reactor startups in order to maintain normal operational primary containment pressure control as The volume of heat loads increase drywell atmosphere temperatures.

gas released during venting operations is expected to be small with respect to purge volumes.

Before purging or venting the containment, airborne contamination levels will be determined and estimates made of expected gaseous activity releases. Selection of release routes and release rates will be made so as to assure compliance with the Technical Specifications.

No special area controls or monitoring procedures are imposed during primary containment purging or venting operations.

5.2.3.8 Primary Containment Atmospheric Control System The Primary Containment Atmospheric Control System (PCACS) has been provided in the design to introduce makeup air or nitrogen into the '

primary containment.

The capability to operate the primary containment with an inert atmosphere has been provided in the design in accordance with previous licensing commitments. This system is capable of reducing and maintaining the oxygen content of the atmosphere and coinplies with the requirements set forth by the American Gas Association. The PCACS will be isolated from the primary containment in the event of an accident.

Basically, the equipment in the PCACS performs two functions: (1) initial purging of the primary containment, and (2) providing an automatic supply of makeup gas. If the inerting system is used, the purging equipment converts liquid nitrogen into gaseous nitrogen.

i Gaseous nitrogen can be introduced into the suppression chamber or the drywell. The PCACS is also capable of automatically providing makeup gas to the primary containment.

5.2.3.9 Drywell Temperature and Pressure Indication Drywell temperature and pressure are recorded in the main control room. These instruments can be utilized to monitor the essential drywell parameters that are used in the Station Safety Analysis in Section 14.

9

' 5.2-17 i

l

PNPS-FSAR 5.2.3.10 . Pressure Suppression Pool Temperature and Level Indication Pressure suppression pool local and bulk temperature is indicated, recorded and alarmed in the main control room. Pressure suppression pool level is continuously indicated in the main control room.

Pressure suppression pool temperature and level can be monitored locally- at the Alternate Shutdown Panel C165. These instruments can be utilized to monitor the essential pressure suppression pool parameters that are assumed for initial values in the Station Safety Analysis in Section 14 5.2.3.11 Drywell I.evel and Torus Pressure The Drywell Level and Torus Bottom Pressure are indicated in the Main Control Room. The level indicator measures from plant elevation 47 feet to the containment purge and vent line, elevation 77 feet. The pressure indicator measures from 0 to 100 psig. These parameters are also provided to the EPIC computer. The parameters are used in conjunction with the Emergency Operating Procedures.

5.2.4 Safety Evaluation

! 5.2.4.1 General The primary containment and its associated safeguard systems l

accomplish the following safety design bases:

1. Acccmmodate the transient pressures and temperatures a

associated with the postulated equipment failures within the

,' containment (safety design basis 1)

, 2. Provide a margin for the effects of a metal water and other chemical reactions subsequent to postulated accidents involving loss of coolant (safety design basis 2)

3. Provide a high integrity barrier against leakage of any

! fission products associated with these equipment failures (safety design basis 3 and 5) l

4. Provide for long term core flooding (safety design basis 4)

5. Provide for rapid actuation of the containment barrier (safety design basis 6)

6. . Store water for the CSCS (safety design basis 7)

7. Maintain the containment parameters during planned operation l

to within those assumed in the Station Safety Analysis (safety design basis 8)

! These factors are considered in the following evaluation of the integrated PCS.

5.2-18 Revisivn 11'- luly 1990

PNPS-FSAR l

. .5.2.4.2 Primary Containment Characteristics Following a Design Basis Accident I

In order to establish a design basis for the pressure suppression containment with regard to pressure alid temperature rating and steam  !

condensing capability, the maximum rupture size of the Reactor Primary System must be defined. For this design, an instantaneous, l

circumferential rupture with double ended flow of one recirculation l line has been selected as a basis for determining the maximum gross '

drywell pressure, and the condensing capability of the pressure suppression system. The selection of a failure of this size for the design basis is entirely arbitrary, since the circumferential failure -

of a recirculation pipe of this magnitude is considered to be of 5.2-1Sa Revisior. 10 - July 1959 l

.)

1 l

PNPS-FSAR

'Although it has been concluded that with the application of conservative I piping design and proven engineering practices pipes will not break in l such a manner as to bring about movement of pipes sufficient to damage the primary containment vessel, the design of the containment and piping l systems does consider the possibility of missiles being generated from )

the failure of flanged joints such as valve bonnets, valve stems, recirculation pumps, and from instrumentation such as thermowells.

The most positive manner to achieve missile protection is through basic station arrangement such that, if failure should occur, the direction of flight of the missile is ~away from the containment vessel. The arrangement of station components takes this possibility into account even though such missiles may not have enough energy to penetrate the containment.

Spatial. separation and utilization of the biological shield to the maximum extent practical are the measures taken .to minimize the possibility of a single potential missile causing a loss of more than one redundant subsection of a vital safety system or a loss of more than ,

one functionally independent safety system.

In order to minimize post-accident containment leakage, the containment penetrations are designed to retain their integrity during postulated accidents.

It is concluded that safety design basis 3 is met.

5.2.4.6 Containment Isolation One of the basic purposes of the PCS is to provide a minimum of one protective barrier between the reactor core and the environmental surroundings subsequent to an accident involving failure of the piping components of the Reactor Primary System. To fulfill its role as a barrier, the primary containment is designed to remain intact before, during, and subsequent to any LOCA in a process system either inside or outside the primary containment. The process system and the primary containment are considered as separate systems, but where process lines penetrate the containment, the penetration design achieves the same integrity as the primary containment structure itself. The process line isolation valves are designed to achieve the containment function inside the process lines when required.

Since a rupture of a large line penetrating the containment and connecting to the Reactor Coolant System may be postulated to take place at the containment boundary, the isolation valve for that line is required to be located within the containment. This inboard valve in each line is required to be closed automatically on various indications of reactor coolant loss. A certain degree of additional reliability is added if a second valve,. located outboard of the containment and as close as practical to it, is included. This second valve also closes 1

automatically on an isolation signal. Both valves shall receive the isolation (closure) signal even if normally closed during reactor 5.2-21 ' Revision 16 - June 1994

1 1

I PNPS-FSAR operation. If _ a failure involves one valve, the second valve -is

available to function as the containment barrier.

By. physically separating the two valves there is less likelihood that a ,

failure of one valve would cause a failure - of the second. The two l

. valves in series are provided with independent power sources.

The ability of the. steam line penetration and the associated steam line isolation valves to fulfill the containment objectives under several postu} ated break locatibns in the steam line is described below, and demonstrates the adequacy of the isolation valve design:

1. The failure occurs within the drywell upstream of the inner isolation valve Steam from the reactor is released into the drywell and the resulting sequence is similar to that of a design basis LOCA except that the pressure transient is less severe since'the blowdown rate is slower. Both isolation valves close upon receipt of the signal indicating low water level in the l reactor vessel. ?!iis action provides two barriers within the steam pipe passing through the penetration and prevents further flow of steam to the turbine. Thus when the two isolation valves close subsequent to this postulated failure, the primary containment barrier is established, and '

the reactor is effectively isolated from the external environment

2. The failure occurs within the drywell and renders the inner isolation valve inoperable Again the reactor steam will blow down into the primary containment. The outer isolation valve will close upon receipt of the low water level signal, establishing the l primary containment barrier

3. The failure occurs downstream of the inner isolation valve either within the drywell or within the guard pipe

' Both isolation valves will close upon receipt of a signal l indicating low water level in the reactor vessel. The guard pipe is designed to accommodate such a failure without damage to the drywell pet.etration bellows , and the design of the pipeline supports protect its welded juncture to the drywell vessel. Thus the reactor vessel is isolated by the closure of the inner isolation valve and the primary containment barrier is established by closure of the outer isolation valve. It should be noted that this condition provides two barriers between the reactor core and the external environment

4. The failure occurs outside the primary containment between the guard pipe and the outer isolation valve 5.2 22 Revision 16 - June 1994 8

PNPS-PSAR 1

(

The steam will blow directly into the pipa tev41 through-the blowout panel and into the Turbine Building until the isolation valves are automatically closed. Closure of the inner isolation valve places a barrier between the reactor.

This. barrier serves to core and the external environment. integrity.

isolate the reactor and complete the containment Closure of the outer isolation valve in this . incident serves no useful purpose

5. The failure occurs outside the primary containment anu renders the outer isolation valve inoperative The primary containment barrier and isolation of the reactor is achieved by closure of the inner isolation valve

6. The failure occurs outside the primary containment between the outer isolation valve and the turbine The steam will blow down directly into the pipe tunnel or the Turbine Building until botn isolation valves' are automatically closed. This action isolates the reactor, establishes the primary containment barrier, and places two barriers in series between the reactor core and the outside environment The exceptions to the arrangement of isolation valves described above or Reactor Primary for lines connecting directly to the containmentto a less desirable i System are made only in cases where it leads system situation because of required operation or cases In the maintenance where, offorthe example, in which the valves are located. special the two isolation valves are located outside the containment, attention is given to assure that the piping to the isolation valves has an integrity at least equal to the containment. ]

i When the TIP f

The TIP system isolation valves are normally closed. l system cable is inserted, the valve of the selected tube opens Insertion. l automatically and the chamber and cable are inserted. l calibration, and retraction of the chamber and cable requires '

approximately 5 min. Retraction requires approximately 1 1/2 min.

If closure of the valve is required during calibration, the isolation signal causes the cable to be retracted and the valve to close 4 automatically on completion of cable withdrawal.

It is neither necessary nor desirable that every isolation For exampte,valveif close simultaneously with a common isolation signal. it would be important a process pipe were to rupture in the drywell, to close all lines which are open to the drywell, and some effluent However, under these process lini.e such as the main steam lines.  !

conditions, it is essential that containment and core cooling systems be operable. For this reason, specific signals are utilized for isolation of the various process and safeguard systems.

Isolation valves must be closed before significant amounts of fiss2on products are released from the reactor core under DBA conditions.

5.2-23

PNFS-FSAR Because small, a sufficient the amountlimitation of radioactive of fission materials in the will product release recetor be coolant is the coolant accomplished if the isolation valves are closed before drops below the top of the core.

It is concluded that uafety design basis 6 is met.

5.2.4.7 Containment Flooding As is discussed in Section 12, the PCS is designed for the conditions associated with flooding the containment.

It is concluded that safety design basis 4 is met.

5.2.4.8 Pressure Suppression Pool Water Storage the Based upon the Station Safety Analysis presented in Section 14.

stored in the suppression pool is sufficient to As quantity of watercondense the steam from a DBA and to provide water for the CSCS.

discusred in Section 12, the suppression pool is considered in the loading conditions on the PCS.

It is concluded that safety design basis 7 is met.

5.2.4.9 Limitations During Planned Operations 5.2.3.6, 5.2.3.7, and 5.2.3.8, the POS is As is discussed in Sections designed to be kept within the limits of parameters assumed in the

( Station Safety Analysis presented in Section 14 during planned operations.

It is concluded that safety design basis 8 is met.

5.2.4.10 Primary Containment Steam Quenching is designed to contain a pool of The water suppression in order tochamber, or torus, suppress the pressure during aThepostulated LOCA by the Reactor Pr: mary System.

condensing the steam released frem reactor system energy released by re12ef valve operation during operating transients also is released into the suppression pool.

As a result of concerns regarding potential instability of steam condensation in a hot suppression pool, the United States Nuwlear Regulatory Commission (NRO) has imposed pool temperature 12mits for (SRV) operation plant transients involving safety / relief valve (Reference 1). The limits which ensure smooth steam condensation for discharge through quenchers are:

1. For all plant transients involving SRV operation during which the steam flux through the quencher perforations exceeds 94 lbm/ft-

-sec, the suppression pool local temperature shall not exceed 200'F.

5.2-24 Revision 5 - July 1985 j

FNPS-FSAR

2. For all plant transients during which the steam flux through the 42 lbm/ft #-sec, the quencher perforations is less than suppression pool local temperature should be at least 20*F subcooled. This correspends to a local temperature limit of 201.4'T for PNPS.

For plant transients involving SRV operation during which the

3. the quencher perforations exceeds 42 lbm/ft' steam flux through

-see but is less than 94 lbm/ft*-sec, the suppression pool local temperature can be estaclished by linearly interpolating the local temperatures established under items (1) and (2) above.

These limits are depicted in Figure 5.2-19.

An analysis was' done (Reference 2) to show that PNPS complies with the NRC criteria.

Seven transient events have been identified, one of which is exp'ected to result in the maximum long-term suppression pool temperature. The seven events are as follows:

1A. Stuck-open SRV during power operation with one RER loop available.

1B. Stuck-open SRV during power operation assuming reactor isolation due to MSIV closure.

2A. Isolation / scram and manual depressurization with one RRR loop available.

2B. Isolation / scram and manual depressurization with the failure of an SRV to reclose (SCRV).

2C. Isolation / scram and manual depressurization with two RRR This cr.d e demonstrates the pool loops available.

temperature responses when an isolation / scram event occurs under normal power operation (i.e., when all- systems are operating in normal mode).

3A. Small-break accident (SBA) with manual depressurization:

accident mode with one RHR loop available.

3B. Small-break accident (SBA) with manual depressurization and failure of the shutdown cooling system.

The analysis indicated that the maximum temperature occurs during Case 2A. The maximum local pool temperature for this case is 199'T (reference 4) which is less than the 201.4*F limit applicable for low steam flux conditions.

5. -24a Revision 5 - July 1955

PNPS-FSAR l 5.2.4.11 Steam Bypass-Following a reactor coolant pipe break inside the containment, the potential exists for the air-steam mixture within the drywell to pass through various leakage paths into the suppression chamber, thereby causing pressurization of the suppression chamber. This increased back pressure in the suppression chamber might lead to an increase in pressurization of the drywell, and possible overpressurization of the containment beyond the design limits.

r The bypass area is expressed in terms of A/sk, where A is the total The bypass (leakage) area and k isthat thecould pressure loss coefficient.

exist between the drywell maximum allowable leakage area and the suppression chamber is a function of the area of the break as well as the duration of pressurization. The former depends on the P between the dryvell and suppression pool, and the latter relates to the time delay until containment sprays are initiated.

In order to assess this relationship, an analysis was performed with various steam break sizes. For large breaks the AP is high. btf: has Primary a short duration. The ' maximum AP results from the DBA, system breaks greater than approximately 0.3 ft* will result in Figure 5.2-22

rapid depressurization of the primary system.

e shows ' the allowable bypass capacity (A/sk) as a function of primary system break area.

I The allowable A/sk is determined on the basis of the allowable steam mass that can be bypassed without exceeding the containment design pressure of 62 psig. For the Pilgrim Nuclear Power Station the Typically, maximum allowable bypass capacity is an A/s'k =0.13 f t * .

the geometric loss factor would be 3 or greater. Thus, the actual j

E 2

5.2-25 Refiston 5 - July 19*5

- am,

PNPS-FSAR This is allowable bypass area would be approximately 0.2 ft'.

equivalent to a 6 in orifice.

When - calculating . the allowable leakage capacities shown on of events is assumed.

5.2-22, the following sequence Figure Immediately af ter a break in the primary system, a rapid the noncondensible gasesrise in in the containment pressure would occur as For' the drywell are transferred to the suppression chamber.

allowable leakage calculations, no operator action Further, is assumed a 10until min the suppression chamber pressure reaches 25 psig.

delay is assumed before any action is taken to terminate the transient. In addition to the 10 min operator delay, a 5 min delay

'is assumed for corrective action to become effective.

The following . assumptions were made in calculating the allowable leakage capacities:

1. Flow through the postulated leakage is pure steam. This is a conservative assumption as the amount of steam rereased -

into the suppression pool is maximized

2. There is no condensing of the leakage flow on either the suppression ' pool surface or the torus and vent system structure.

This assumption results in a conservative peak pressure calculation Station emergency procedures ensure that operator corrective action appropriate to the postulated events is taken. If the low-low water level point has .not been reached, the operators can depressurize the reactor vessel through the main steam lines to the main condenser or alternately, utilize the relief valves to rapidly reduce reactor pressure. Existing emergency procedures require the initiation of spray mode of the RHKS after the pressure suppression pool level is adequate.

verification that the reactor vessel water Further, the procedures require the initiation of the drywell spray mode of the RHRS if the drywell pressure rises to 10 psig.

5.2.5 Inspection and Testing The following discussion details the surveillance and testing that will be conducted on the various systems or components of the primary containment during construction or station operation.

5.2.5.1 Primary Containment Integrity and Leaktightness Fabrication procedures, nondestructive testing, and sarple coupon tests were made in accordance with the ASMIThe Code of Boilers integrity of and the Pressure Vessels,Section III, Subsection B. The Primary Containment System was verified during construction.

verification included a pneumatic test of the drywell and suppress:on chamber at 1.25 times their design pressure in accordance with code l

requirements.

t l

5.2-26 i

PNPS-FSAR After complete installation of all. penetrations in the drywell and suppression chamber, the vessel was pressurized to the calculated peak accident pressure, and measurements taken to verify that the integrated leakage rate from the vessel did not exceed the maximum allowable leak rate. A second test was run at reduced pressure to containment establish a relationship between leakage rate and pressure. The necessary instrumentation is installed in the station to provide the data required to calculate and verify the leakage rate.

Provisions- are made so that integrated, containment leakage rate tests may be periodically performed during periods of reactor shutdown, in compliance with 10CFR50, Appendix J. Primary Containment Leakage. Testing for Hater Cooled Power Recctors.

5.2.5.2 Penetrations The design permits the testing of penetrations which have resilient seals or expansion bellows without pressurizing the entire containment system. Leak detection may then be accomplished either by the use of soap suds, pressure decay techniques, or other acceptable methods.

Pipe penetrations which must accommodate thermal movement are -

provided with two ply bellows expansion joints. These two ply bellows are provided with test taps so that the space between the plies.can be pressurized to the calculated peak accident pressure to j permit testing of the individual penetrations for leakage. ,

Electrical penetrations are also separately testable. The test taps are located so that the tests of the electrical penetrations can be l conducted without entering or pressurizing the drywell or suppression chamber.

All containment closures which are fitted with resilientThe seals or covers l

gaskets are separately testable to verify leaktightness. l I

on flanged closures, such as the equipment access hatches, the i drywell head, access r aholes, and personnel air lock doors, are provided with double seals, and with a test tap which allows pressurization of the space between the seals without pressurizing l the entire containment system.

)

f 5.2.5.3 Isolation Valves The test capabilities which are incorporated in the PCS to permit leak detection testing of containment isolation valves are separated l

, into two categories. l f

i The first category consists of those pipelines which open into the

- containment atmosphere and do not terminate in closed loops outside the containment, and contain two isolation valves in series. Test I taps are provided between the two valves which permit leakage l monitoring of the first valve when the containment is pressurized.

5.2-27 Revision 12 - Jan 1991 f

i

PNPS-FSAR The test tap can also be used to pressurize between the two valves to permit leakage testing of both valves simultaneously.

The second category consists of those pipelines which connect to the Reactor System and contain two isolation valves in series. A leakoff line is provided between the two valves, and a drain'line is provided downstream of the outboard valve. This arrangement permits monitoring of leakage on the inboard and outboard valves during Reactor System hydrostatic tests, which can be conducted at pressures up to the reactor system operating pressure of 1,000 psig.

Generally, leakage testing is not required for isolation valves contained in pipelines whose terminal end will remain submerged in the suppression pool throughout the duration of the design basis LOCA.

Therefore, these valves are not relied upon to prevent the release of fission products, and therefore do not perform a containment isolation function. Reference Number 5, Section 5.2.9 provides a list of valves which fit the above category.

Isolation valve closing time is determined during the functional performance test performed prior to reactor startup.

5.2.6 Nuclear Safety Requirements for Plant Operation l The entries in this section represent the nuclear safety requirements l for the PCS for each BHR operating state which represents an extension of the stationwide BWR systems analysis of Appendix G. The following

' referenced portions of the safety analysis report provide information justifying the entries in this section:

1 Reference Information Provided

1. Preceding portions of Description of PCS Section 5.2

2. Section'7.2, Description of PCICS Primary Contain-ment and Reactor Vessel Isolation Control System

3. Station Safety Anal- Analyses verifying primary ysis, Section 14 containment responses and radiological effects of pos-tulated accidents

4. Station Nuclear Identification of condi-Safety Operational tions and events for which Analysis, Appendix G PCS is required

5. Bodega Bay Prelim- Pressure suppression test inary Hazards Sum- information mary Report, Appen-dix 1 Docket 50-205 5.2-28 Revision 12 - Jan 1991

PNPS-FSAR l

6. Jacobs, I.M., Guide- Describes methods used to lines for Determining establish allowable repair Safe Test Intervals times and Repair Times for  ;

Engineered Safeguards.

General Electric Co., l

)

l Atomic Power Equipment l

Department, APED-5736, '

April 1969 Each detailed requirement in the following analysis is referenced, if possible, to the most significant station condition originating a need

' for the requirements by identifying a matrix block on one of the Matrix 3 sheets of Table G.5-3. The matrix block referenced is given in parentheses -beneath the detailed requirements in the " minimum t required for action" section.

The matrix block references identify the BHR operating state, the event number, and the system number. For example. F39-82, identifies BWR operation state F (Matrix 3), event (row) No. 39, and system t

! (column) No. 82.

Minimum Required for Action i 1. Vacuum Relief System l (C39-109) (E39-109)

(039-109) (F39-109)

2. Primary Containment (C39-82) (E39-82)

(D39-82) (F39-82) 4 3. Drywell Pressure and Temperature Indicators (C2-103) (E2-103)

(D3-103) (F4-103)

4. Pressure Suppression Pool Hater Level and Temperature Indicators (A35-104) (D3-104) (F4-104)

(B35-104) (D39-104) (F39-104)

(C2-104) (E2-104)

(C39-104) (E39-104)

5. Pressure Suppression Pool Hater Storage (A35-83) (D39-83)

(B35-83) (E39-83)

(C39-83) (F39-83)

I 5.2-29 Revision 12 - Jan 1991

I PNPS-FSAR j l

l Requirements are placed on the operating status of systems essential to containment to assure their availability to control the release of l any radioactive material from irradiated fuel in the event of an  !

accident condition. The PCS provides a barrier against uncontrolled i release of fission products to the environs in the event of a break in the Reactor Coolant. Systems. Whenever the reactor is in states C, D, E, and F (with nuclear system pressurized), failure of the Reactor Coolant System could cause rapid expulsion of the coolunt from .the  :

l reactor, with an associated pressure rise in'the primary containment.

Primary containment is required, therefore, to limit the release of fission products to the station environs so that offsite doses would l be well below the values specified in 10CFR100.

The calculated radiological doses given in Section 14 were based on an assumed. leakage rate of 0.5 percent. Increasing the assumed leakage rate at 56 psig' to 2.0 percent would increase those doses by [

approximately a factor of four, still leaving a substantial margin between the calculated dose and the 10CFR100 regulation.

The suppression pool water volume provides the heat sink for the Reactor Coolant System energy released following the LOCA. In states A and B the suppression pool water is available as a source of makeup water to replace possible leakage from the reactor vessel and primary

• system.

The maximum water volume limit allows for an operating range without significantly affecting the accident analyses with respect to free air volume in the suppression pool. The maximum pool bulk temperature of 130*F would accommodate a complete accident blowdown with minimum water volume without exceeding the design temperature limit of 170*F ,

i immediately af ter blowdown. The design minimum water temperature of 40*F assures that the water is always in the liquid state. ,

Suppression pool temperature limits have been selected to assure

reactor depressurization can occur without high pressure suppression i chamber loadings caused by instability during steam condensation.

! The Drywell Suppression Pool Vacuum Breaker System is required to prevent water oscillation in the downcomers due to low steam flow

. rates in the downcomers, and to provide protection against negative pressure conditions in the containment vessel. Allowing one valve to be inoperative reduces the total vacuum relief area by only 10 l percent. Since the valves are totally enclosed within the containment, possible leakage through them does not affect the containment system leakage.

[ The Suppression Pool Reactor Building Vacuum Relief System assures that the primary containment is not operated at a significant negative pressure relative to its surroundings. The 0.5 psi differential

! pressure setting was chosen on the basis of Relief System pressure drop, valve opening times, and peak mass flow to limit the external pressure on the suppression chamber to less than its design value of I

5.2-30 Revision 12 - Jan 1991 l

l

PNPS-FSAR 2.0 psig. The Vacuum Relief System is a redundant system and full relief capacity is available through either valve. If one vacuum breaker or its block valve becomes inoperable, there is no immediate threat to primary containment integrity, thus, reactor operation may continue while repairs are being made, provided the repair procedure does not violate primary containment integrity. Possible leakage of these valves is included in the containment system integrated leakage

rate tests performed periodically.

5.2.7 Current Technical Specifications l The current limiting conditions for operation, surveillance requirements, and their bases are contained in the Technical Specifications referenced in Appendix B.

5.2.8 Pipe Break Transient Analysis 5.2.8.1 Pipe Mechanical failure and Safety Design The Pilgrim Nuclear Power Station primary containment satisfies safety design basis 1 by its capability "to accommodate the transient pressures and temperatures associated with the postulated equipment failures within the containment." The intent of safety design basis 1 is to provide a basis for determining the primary containment internal design pressure and associated temperature, and that basis is that the primary containment must remain functional after accommodating the largest mass flow and energy release associated with the design basis LOCA. See Section 5.2.4.1.

Section 5.2.4.2 discusses the selection of The the failure conditions capability that to satisfy establish containment design parameters.

safety design basis 1 is demonstrated in the primary containment response analysis to the design basis LOCA present in the Station Safety Analysis in Section 14. It is not the intent, nor has it ever been assumed that it would be the intent, that safety design basis 1 be used as a basis for evaluating the abilities of the primary containment to accommodate the mechanical forces and energies that might be associated with the movement of unrestrained pipe during a postulated LOCA.

l l

5.2-31 Revision 12 - Jan 1991 l

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1 PNPS-FSAR INTENTIONALLY LEFT BLANK 5.2-32 Revision 12 - Jan 1991

PNPS-FSAR INTENTIONALLY LEFT BLANK 5.2-33 Revision 12 - Jan 1991 l

J

i PNPS-FSAR i

e Section 5.2.4.5 discusses containment integrity protection within the scope and the intent of safety design basis 3 and defines the i pertinent loading considerations that have been evaluated in order to meet' safety design basis 3.

.In order to minimize the probability of an instantaneous failure in the Reactor Primary System piping, design provisions were made to i

minimize or identify conditions that could lead to such a failure.. ,

As discussed in Appendices J.2.4 and F.2.6, the Reactor Primary System is designed to meet the intent 'of Criterion 35 of the Proposed i AEC General Design Criteria, thus reducing even _further the extremely low probability of an instantaneous piping- failure due to brittle fracture, j In addition, a Nuclear System Leak Detection System, as described in

- Section 4.10, is provided to identify primary system leakage rates

well below those leakage rates which correspond to the critical size

for rapid crack propagation. The capability of this system to i

identify these leakage rates will provide significant protection

~ against an instantaneous primary system piping failure due to crack

- propagation by allowing station personnel sufficient time to take appropriate corrective measures. Supplementary protection will be j provided by the comprehensive inservice inspection program discussed l

in Appendix K.

1 l In conclusion, the design fabrication, testing, and inspection of the

Reactor Primary System has emphasized the elimination of potential ,

i causes for instantaneous piping failures, and thus obviates the need

to design the PCS to withstand the mechanical effects of the fai' led j pipe. Therefore, the protection of the primary containment from the mechanical effects of an unrestrained failed pipe is not a safety

design basis for Pilgrim Nuclear Power Station.

5.2.8.2 Pipe Protection System

! As a result of an investigation, selected areas of the interior of the drywell shell were protected to reduce the possibility of breaching of the primary containment by postulated failure of a large, unrestrained pipe in the primary pressure boundary.

All pipe penetrations through the drywell have been designed to withstand the forces and moments resulting from a pipe rupture inside the drywell. Main steam and feedwater piping have restraints outside the drywell for protection of the penetration assemblies and outboard isolation valves.

The piping systems considered ~ for postulated failure and having the potential to breach the containment are those which are located within the spherical section of the drywell and normally pressurized to reactor pressure (main steam, HPCI steam supply, feedwater, RHR).

These large pipes are postulated to fail at circumferential butt welds with the jet reaction force acting normal to the rupture surface, and resulting in pipe rotation around a plastic hinge. The 5.2-34 Revision 12 - Jan 1991

l PNPS-FSAR-

{

drywell areas requiring protection are shown' on Figure 5.2-23, and are those areas in the spherical section where the single postulated l

large pipe weld failure could result ir, pipe movement to the extent l'

that the ruptured pipe could contact the interior of the drywell shell with sufficient energy to perforate the drywell.  ;

! The protection system consists generally of steel members attached to  :

o a reinforcing plate. The protection system is arranged to receive a

' postulated rupture pipa, absorb a portion of the impact energy, and s distribute the impact load over an area of the drywell shell such that the combined energy absorbtion capacity of the protection system and the drywell shell is greater than the impact energy of the '

ruptured pipe. The protection system and the drywell shell will deform through the 2 in air gap between the drywell and the concrete ,

i shield without causing breaching of the drywell. Details of the protection system are shown on Figure 5.2-23.

/

1 Areas of the. spherical section of the drywell- shell requiring protection have been determined by plotting the potential area at '

which each ruptured pipe end could contact the drywell when the pipe

' is rotated around various possible plastic hinge points. The force l

causing pipe movement and deformation around a plastic hinge is the jet reaction resulting from blowdown from the reactor system.

1 The impact energy of a ruptured pipe has been determined as a i function of the jet reaction force, pipe plastic bending moment, and the configuration of the pipe with respect to the drywell shell. The

energy required for perforation of the drywell shell has been l

determined from an empirical relation developed from a series of E experiments using steel projectiles.

l The protection system component size and placement is based upon the

requirement to distribute the pipe impact energy over a sufficient l area of the drywell shell, such that the combined energy absorbing capability of the protection system and drywell shell is greater '

the impact energy of the ruptured pipe. The steel beams are arra;. M to minimize the possibility of a ruptured pipe end from resulting in

]

a localized load bearing directly on the drywell shell. The beams

are attached to a steel plate located between the beams and the drywell shell. This plate results in increasing the energy absorbing

_ capability of the drywell by increasing the impact area, and increasing the effective thickness of the drywell shell. The plate also serves as a means of restraining the beams against potential jet impingement loads and the component of pipe impact loads tangential to the drywell shell.

The protection system is attached to the drywell shell at the weld '

pads with additional support from the floor structures as required.

The protection system supports are designed to withstand the loads from the Safe Shutdown Earthquake.

The protection system components have been selected and located such l' that maximum protection is provided for the dryvell : hell against i

5.2-35 I

-, - . - .- - -_ .=. - . -.

i PNPS-FSAR i i i

postulated pipe ruptures with minimum interference to required access

.for-inservice inspection.

Pipe ruptures within the cylindrical section of the drywell have been l considered and no protection is required because:

Pipe movement distances to contact the drywell are 1.

insufficient to obtain an impact energy exceeding the energy required to perforate the 1 1/4 in shell thickness i 2. The close proximity of the drywell shell to the piping systems is such that pipe rotation around a plastic hinge is insufficient to result in the ruptured end becoming a localized load on the drywell The analysis and basis for design of the protection system is conservative because:

! 1. Jet reaction forces have not been reduced due to the throttling effect of partial pipe closure at the plastic hinge point

2. Pipe impact energies have not been reduced by the energy absorbed by pipe deformation at the point of contact between the protection system and the drywell shell

3. Impact energy required to perforate the drywell shell is based on test data using tool steel projectiles, and is therefore lower than the energies required for perforation with typical pipe materials

5.2.8.3 Design Basis Line Break The design basis steam line break accident (SLBA) is described in detail in Section 14. This accident is assumed to result in a cogplete guillotineAllbreak otherofbreaks the main steam attached in piping line, resulting to theinvessel a 1.74 ft break area.

i above the core result in peak clad temperatures which are lower than those resulting from the SLBA. Since there are no perforations for the SLBA, there will be none for smaller steam line breaks. While the SLBA evaluated in Section 14 considers isolation of the reactor i

vessel, an analysis of an SLBA inside the primary containment (i.e.,

no isolation) is described in Section 5.2.3.2. The results of this analysis show that the core will remain covered throughout the

- transient and the resultant peak clad temperature will be less than normal operating temperatures, which are well below the temperature where clad perforation could occur. As in the case of the design basis SLBA, all other smaller steam lines which could fail in such a manner that isolation is not achieved would also not result in clad perforations. Consideration of those liquid breaks which could conceivably result in containment breaching as a result of pipe whip has also resulted in the conclusion that no fuel perforations will occur.2)In 0.5 ft particular, incore for thealso is feedwater not line uncovered.

break (approximately It can 5.2-36 Revision 14 - June 1992

PNPS-FSAR 4

^;

therefore be concluded that the resultant radiological exposures for the above pipe failures will at the maximum be based on only that

activity contained in the primary coolant, which is discharged to the secondary containment.

1 i To provide an upper limit to the radiological -exposures, the

j. assumptions have been made that: .

I 1. - All of the primary coolant which contained activity is .

) eventually discharged to the secondary containment

? 2. .Considering the thermodynamics of the coolant discharged, a i maximum of 1/3 of the coolant is flashed to steam resulting 2 in the release to the secondary containment of 1/3 of the coolant activity i

3. Consideration of the condensing and plateout surfaces that i the released steam will have to come in contact with prior to i being released from the top _ of the Reactor Building results

in a minimum reduction factor of 3 for the released iodine f activity

4. The ~ activity is released from the top of the Reactor Building under those meteorological conditions, which maximize the I offsite exposures

! 5. The activity contained in the reactor coolant is consistent I with an offgas emission rate of 105microcuries/sec Based on the above considerations, the resultant site boundary thyroid l dose is 0.08 rem while the LPZ thyroid dose is 0.002 rem. If the conservative assumption is made that downwash of the released effluent

occurs and that the coolant activity is at a level consistent with the i technical specification offgas activity (i.e., 0.9 ci/sec), the 1 resultant site boundary thyroid dose is 15 rem and the LPZ thyroid l

dose is 0.6 rem, both of which are well below the 300 rem guideline l set forth in 10CFR100.

J 5.2.9 References I 1. Bodega Bay Preliminary Hazards Report, Appendix I, Docket 50-205, f December 28, 1962.

I 2. General Electric Company. "PNPS Unit 1 Suppression Temperature Response," NEDC-22089-P, March 1982.

Pool

! 3. J. M. Caroll, BECo Letters to NRC, May 15, 1973.

! 4. General Electric Company, " Pilgrim Suppression Pool Temperature i Analysis," Letters from R. Thibault to G. McHugh, December 1982.

! 5. Franklin Research Center Technical Evaluation Report, " Containment Leakage Rate Testing" TER-C5257-40, May 5, 1981.

! 5.2-37 Revision 9 - July 1988 i

PNPS-FSAR

6. NRC letter, " Licensee Response to IE Bulletin 79-08 and Acceptability of Single Check Valves as Containment Isolation for Pilgrim," Ronald Eaton (NRC) to G. H. Davis, February 4, 1991.

Electric Company, "Drywell

7. E. H. Hoffman. et. al., General Temperature Analysis for Pilgrim Nuclear Power Station,"

EAS-98-0887; August, 1987.

~

8. Letter, GE to BECo, " Safety Evaluation of Proposed Capping of Certain Drywell Spray Nozzles," G-HK-7-157, dated April 20, 1987.

3

)

i 1

'l i

4 J

t 5.2-38 Revision 14 - June 1992

k PNPS-FSAR TABLE 5.2-1 ,

PRIMARY CONTAINMENT SYSTEM

, PRINCIPAL DESIGN PARAMETERS AND CHARACTERISTICS 4

Pressure suppression chamber: ,

Internal design pressure +56 psig External design pressure +2 psig Drywell:-

! Internal design pressure +56 psig External design pressure +2 psig Drywell-free volume . . . . . . . . . . . . . . (approx) 147,000 ft3 Pressure suppression chamber free volume . . . . . . . . . . . . . . . . . . . . . (approx). 120,000 ft3 Pressure suppression pool water volume, maximum . . . . . . . . . . . . . . . . .(approx) 94,000 ft3 Pressure suppression pool water volume, minimum . . . . . . . . . . . . . . . . .(approx)-84,000 fta Submergence of vent pipe below pressure (approx) 3.00 to 3.25 ft supprecsion pool surface . . . . . . . . . .

Design temperature of drywell . . . . . . . . . . . . . . . . 281*F Design temperature of pressure suppression chamber . . . . . . 281*F j

Downcomer vent pressure loss factor . . . . . . . . . . . . . . 6.21 Break area / total vent area . . . . . . . . . . . . . . . . . . 0.0194  ;

1 Drywell free volume / pressure suppression chamber free volume . . 1.34 ,

l Primary system volume / pressure suppression pool volume . . . . 0.268 Drywell free volume / primary system volume . . . . . . . . . . . 7.4 Calculated maximum pressure during blowdown:

Drywell . . . . . . . . . . . . . . . . . . . . . . . . . 45 psig Pressure suppression chamber . . . . . . . . . . . . . . . 27 psig Initial pressure suppression chamber temperature rise . . . . 35*F J 1 of 1 Revision 2 - July 1983

PNPS-FSAR TABLE 5.2-2 >

DRYWELL ATMOSPHERE COOLING DATA SHEET

'- Location Average' . Maximum

• General 135'F 148'F Recirculation Pump -

Motor Area

- 128*F Entering Air, Temperature to Cooling Units 135'F 148'F

~ Leaving Air. Temperature-from Cooling Units 85*F 95*F Cooling Water Supply Temperature 75'F 85 F Cooling Water Return Temperature 90 F -100'F Drywell Heat Gain 2.4 X 108-Btu /hr 3.4 X 108 Btu /hr Total Cooling Unit Capacity 3.6 X 108 Btu /hr 5.6 X 108 Btu /hr Total Cooling Unit Fan 72,000 ft 3/ min 110,000 ft 3/ min capacity <

Total Fan Brake hp 54.8 67.8 l i

Drywell Temperature 10 hr after shutdown 105'F 105 F NOTE:

• As a result of higher cooling water supply temperature and extra heat load from scram of the control rod drives.

1 of 1

PNPT-FSAR 7 - . ,,

5.3 SECONDARY CONTAINMENT SYSTEM  ; ,

3 e u n w to 5.3.1 Safety Objective reference The safety objective of the Secondary Containment System (SCS), in conjunction with other engineered safeguards and nuclear safety systems, is to . limit the release to the environs of radioactive materials so that offsite doses from a postulated DBA will be below the guideline values of 10CFR100.

5.3.2 Safety Design Basis The safety design bases of the SCS are as follows:

1. The SCS shall be designed to provide secondary containment when the primary containment is operable and when the primary containment is open

2. The SCS is designed with sufficient redundancy so that no single active system component failure can prevent the system from achieving its safety objective

3. The SCS shall be designed in accordance with Class I design criteria. (Exception to this is the containment access Since simultaneous LOCA's and SSE's are not locks.

po'stulateri, the access locks shall be designed in accordance with Class II design criteria. The access lock door lying directly in the SCS shall be designed in accordance with Class I design criteria so that the possibility of a ground level release to the environs through the access locks is eliminated if a seismic event were to follow or precede an accident which results in a contaminated reactor building atmosphere.) The SCS is not designed to withstand tornado loads

4. The secondary containment shall be designed to limit the ground level release to the environs of airborne radioactive

' materials so that offsite doses from a design basis fuel handling, or loss of coolant accident (LOCA) will be below the guideline values stated in 10CFR100

5. The Reactor Building shall be designed to contain a positive internal pressure of at least 7 in of water

6. The SCS shall be designed to be sufficiently leaktight to allow the Standby Gas Treatment System (SGTS) to reduce the Reactor Building pressure to a minimum subatmospheric pressure of 0.25 in of water, under neutral wind conditions, when the SGTS fans are exhausting Reactor Building atmosphere at a maximum of 4,000 ft'/ min

7. The Reactor Building Isolation and Control System (RBICS) shall be. designed to isolate the Reactor Building sufficiently fast to prevent fission products from the postulated fuel handling accident from being released to the 5.3-1 Revision 6 - July 1986

T PNPS-FSAR _l environs through the normal discharge path The SCS is' provided with means .to conduct periodic tests to

~

' 8.

verify system performance-

, l l

l P

5 6

i

[

+

5.3-la Revision 6 - July 1986 i

t' 9'

r r-e.- $ +e-- 9

PNPS-FSAR 5.3.3 Description 5.3.3.1 General The SCS consists of four subsystems. These subsystems are the Reactor Building, the RBICS, the SGTS, and the main stack. The SCS surrounds the Primary Containment System, and is designed to provide secondary containment for the postulated LOCA. The SCS also surrounds the

-refueling facilities and is designed to provide primary containment for the postulated refueling accident.

The SCS utilizes four different . features to mitigate the consequences of a postulated LOCA (pipe break inside the drywell) and the refueling accident (fuel handling accident). The first feature is a ' negative pressure barrier which minimizes the ground level release of fission products by exfiltration. The second feature is a low leakage containment volume which provides a holdup time for fission product decay prior' to release. The third feature is the removal of particulates and iodines by filtration prior to release. The fourth feature is the exhausting of the secondary containment atmosphere through an elevated release point which aids in dispersion of the effluent by atmospheric diffusion. Each of the features is provided by a different combination of subsystems: the first by the Reactor Building, the RBICS, and the SGTS exhaust fans; the second by the Reactor Building and the RBICS; the third by the SGTS filters; and the fourth by the main stack.

5.3.3.2 Reactor Building The Reactor Building completely encloses the reactor and its pressure suppression Primary Containment System. The Reactor Building houses the refueling and reactor servicing equipment, new and spent fuel storage facilities, and other reactor auxiliary and service equipment. Also housed within the Reactor Building are the CSCS, Reactor Cleanup Demineralizer System, Standby Liquid Control System (SLCS), Control Rod Drive (CRD) System, Reactor Protection System 3

(RPS) and elettrical equipment components.

The structural design features of the Reactor Building are described

, in Section 12. Discussions of the Reactor Building's Class I design are included in Section 12 and Appendix C. The Reactor Building is also designed to meet the shielding requirements discussed in Section 12.

5.3.3.3 Reactor Building Isolation and Control System The RSICS serves to trip the Reactor Building supply and exhaust fans,

, isolate the normal ventilation system, and provide the starting signals .for the SGTS in the event of the postulated LOCA inside the drywell, or the postulated fuel handling accident in the Reatter

. Building. Eitner cf tac signals will initiate the SCS. These i signals, which indicate a LO:A inside the drywell, are high drywell pressure or low rea:ter water level. In addition, radiation mor.itors in the operating (refuelin;) floor ventilation exhaust duct, which 5.3-2 4

PNPS-FSAR indicate a fuel handline. accident, can initiate the SCS. Secondary containment can also be iLitiated manually from the control room.

Normally open, air-operated isolation dampers are provided on the discharge side of the Reactor Building and operating floor supply fans. Normally open. air-operated isolated dampers are provided on the intakes to the operating floor ventilation exhaust fans, the clean area exhaust fans, the contaminated area exhaust fans .(upstream of the filter assemblies), and the control rod drive maintenance room exhaust fan. See Figure 5.3-1. Two dampers in series are provided throughout the isolation system to. provide the required redundancy. Both dampers fail closed upon loss of de power to the solenoids or upon loss of instrument air to the dampers. The isolation dampers are piston operated and designed to close after receipt of the secondary containment initiation signal to prevent- release of radioactive material from the secondary containment. The refueling floor exhaust isolation dampers must close in 3 see to isolate the most direct path outside secondary containment. A 5 sec closing time is sufficient for the remaining dampers.

Penetrations of the secondary containment are designed to have leakage characteristics consistent with secondary containment leakage requirements. Electrical penetrations in the Reactor Building are designed to withstand normal environmental conditions and to retain their integrity during the postulated fuel handling accident and the LOCA inside the drywell . Two . interlocked sealed doors on the equipment and personnel ' access locks assure that building access can not interfere with maintaining the secondary containment integrity.

All normally open drains which are open both to the secondary containment and the outside atmosphere are provided with water seals to maintain containment integrity. This is exemplified by the four 14 in dewatering lines for the reactor auxiliary bay floor sumps. These lines penetrate the secondary containment boundary, two below each of the two sumps, and terminate in a pair of troughs within the torus compartment. The two 4 f t cubic shaped troughs, located adjacent to the east wall, maintain containment integrity by providing water seals for each of the four lines. High and low levels are alarmed in the control room. On low level, the operators are directed by procedure to refill the troughs, to ensure containment integrity.

5.3.3.4 Standby Gas Treatment System The SGTS consi:,ts of two, similar, parallel air filtration assemblies I separated by an 18 in thick concrete block wall and completely enclosed within a Class I structure. Each of the filtration assemblies are full capacity. Each consists of a demister (use of the demister is optional), an electrical heating coil, a high efficiency particulate absorber (HEPA), two charcoal filter beds, and a final HEPA filter. With the Reactor Building isolated, each of the two fans has the necessary capacity to reduce and hold the building at a minimum subatmospheric pressure of 0.25 in of water.

5.3-3 Revision 10 - July 1989

PNPS-FSAR ,

I f

L Each fan has a design flow rate of 4,000 ft /3min. Motor-operated exhaust fan outlet damper controls._are provided to maintain the required negative pressure. See Figure 5.2-17. The system _ consists i of two filter banks. Loss of de power and/or supply air to the solenoids causes the valves in filter bank A to open and the valves in filter bank B _ to close. A dedicated air system is provided for long-term damper actuatio'n. The air system consists of two high i

! pressure air tanks, a pressure reducing regulator with inlet and j~ outlet pressure gages, relief valve, two low pressure gages, one low pressure switch, four low pressure air receivers, solenoids valves and i manual valves, and tubing required to make a complete system. The low pressure section of the system is designed to Class I design criteria. The high pressure section is designed to Class II design i

l criteria. The low pressure system. of the dedicated air system is i designed to provide two complete actuation cycles of the SGTS filter isolation dampers / valves and allow for air system. leakage for i long-term operation.

l The demister is designed to remove entrained water droplets and mist from the entering air stream. Use of the demister is optional since there are no design basis accident scenarios which would expose the SGTS to water droplets or mist. The electric heating coil is designed

to reduce -the relative humidity of the air stream to 70 percent. An interlock with its associated 1xhaust fan prevents the heating coil from operating when the fan is shut down. Each HEPA filter is designed to be capable of removing at least 99.97 percent of the 0.30 i micron particles which impinge on the filter. The charcoal filters are iodide-impregnated activated carbon filters capable of removing in excess of 99 percent of the iodine in the air stream with 10 percent i of the iodine in the norm of methyl iodide (CH 3I) under entering )

i- conditions of 70 percent relative humidity.

The accident evaluations using the standard NRC approach are described ,

i in Section 14.9. In these analyses the SGTS charcoal filters were i i credited with removal of 95 percent of the influent iodine. l

! The system will start automatically upon a high radiation signal from

! the operation (refueling) floor ventilation exhaust duct monitor, or 3

upon receipt of high drywell pressure or low reactor water level j

! signals. The system can also be manually started from the control room. Upon receipt of any of the initiation signals. The AUTO "A" train fan and heater start and its associated isolation valves open, i the STANDBY "B" train suction valve opens, starting the "B" train fan j

and heater and in turn, opening ti.e "B" train discharge isolation

valve. In the event that the "A" train is down for maintenance the '

l

STANDBY "B" train suction valve opens when the train mode switch is in the MAINTENANCE position. Each fan and heater draws air from the >

isglated reactor building at a flow rate up to approximately 4000 ft#/ min. After a preset time delay period, the "B" train suction valve closes, which in turn trips the "B" train fan and heater, i closing its discharge valve. Cross-connections between the filter i

trains are provided to maintain the required decay heat removal

! cooling air flow of low humidity air on the charcoal filters in the j inactive treatment train.

5.3-4 Revision 11 - Jcly 1990 i

m -- _ --- - __v_ . , - - -_ _ _ _ _ = _ _. -- - - - - -

PNPS-FSAR The STANDBY "B" filter train is automatically started in the event of train heater, fan, air supply and/or power failure. In an - AUTO "A" addition, the:"B" train is automatically started in the STANDBY mode, whenever the STANDBY "B" train suction valve is not fully CLOSED.

The

' system discharges to the main stack through a 20 in underground line.

The SGTS fans are powered from the emergency service portions of the auxiliary power distribution system.

l

. Drywell and torus purge exhaust can also be directed to the SGTS for L processing before release up the main stack. See Section 5.2. ,

• The High Pressure Coolant Injection System (HPCI) gland seal steam condenser exhauste.- discharge is also routed to the SGTS during '

i accident conditions. . The Reactor Building Heating and Ventilating

< ' System is discussed in Section 10.9.

I During a severe accident, the torus can be directly vented to the main stack bypassing the SGTS. See Section 5.4.7.

I 5.3.3.5 Main Stack The' location of the main stack is shown on Figure 1.6-1. The top of the stack is at elevation 400 ft msl. The structural design of the i stack is discussed in Section 12.

i

- 5.3.4 Safety Evaluation i -

The SCS provides the principal mechanisms for the mitigation of the l

i consequences of an accident in th Reactor Building. The primary and secondary containment act together to provide the principal mechanisms

for the mitigation of the consequences of an accident in the drywell.

If the leakage rate of the building is low, and the leakage air is filtered and discharged to the elevated release point (utilizing the SGTS and the main stack) the offsite radiation doses that result from postulated accidents are reduced significantly. The Reactor Building

.l is a Class I structure (with the exception of the secondary containment access locks which are Class II structures) designed in

~

accordance with all applicable codes. Design oj/ min the Reactor Building j

for a maximum inleakage rate of 4,000 ft at a building i subatmospheric pressure of 0.25 in of water at neutral wind conditions, results in a low exfiltration rate even during high wind 4 conditions. ,

! In the event of a pipe break inside the primary containment or a fuel handling accident, Reactor Building isolation will be effected and the

SGTS will be initiated. Both SGTS exhaust fans will start. After a

preset time delay, one fan is stopped.

With the Reactor Building isolated, each fan in the SGTS has the capability to hold the building at a subatmospheric pressure of 0.25 in ** * " " "I * * " "I " " # #"

• ft /3 min. Exhaust fan' outlet damper controls on each fan are provided 1

to maintain the required flow rate, 5.3-5 Revision 11 - July 1990 l

PNPS-FSAR The RBICS performs the required isolation actions of the SCS following-receipt of .the appropriate initiation signals. Following initiation, the Reactor Building ventilation isolation dampers close within a specified time to prevent release of radioactive material from the secondary containment. The refueling floor exhaust isolation dampers must close in 3 see to isolate the most direct path outside secondary containment. A 5 sec closing time is sufficient for the remaining dampers. The RBICS also automatically trips the Reactor Bu:1 ding supply and exhaust fans, and . starts the SGTS. The normal design flow rate in the geactor Building operating (refueling) floor exhaust duct is 40,000 ft / min. During shutdowns, the flow rate is increased to approximately 50,000 ft3 / min at which time it takes more than 3 sec for fission products released in any postulated fuel handling accident to travel from the: operating (refueling) floor ventilation exhaust radiation monitors to the isolation dampers. Thus, no direct release of fission products to the environment (bypassing the SGTS filtration processes, and the elevated release point provided by the main stack) is possible, except when the direct torus vent path is used following a beyond design basis accident.

I The SGTS filters exhaust air from the Reactor Building and discharges the processed air to the main stack. The system filters particulates and iodines from the air stream in order to reduce the level of airborne contamination released to the environs via the main stack.

When the system is exhausting from the Reactor Building, the building is held at a minimum subatmospheric pressure of 0.25 in of water.

Appendix G identifles requirements for establishing secondary containment (Safety Action 27), following an assumed pipe break nside the primary containment (Event 39), and following an assumed spent fuel handling accident (Event 40). Secondary containment is not l

l I

.5.3-Sa Revision 11 - July 1990

I PNPS-FSAR.

,' established following assumed pipe failures which result in the release of steam into the Reactor Building (Event 41).

[

! The- following piping which. is located within the Reactor Building and l normally contains hot fluids at reactor pressure was considered: High Pressure Coolant Injection (HPCI) turbine steam supply line; Reactor

Core Isolation Cooling (RCIC) turbine steam supply line; and Reactor j Hater Cleanup System (RNCU) piping. ,

i j The maximum rate. of ' steam release into the Reactor Building and the e

corresponding period of- steam release was calculated for the above l piping

! 1. HPCI steam line:

i Maximum release rate, 300 lb/sec of steam

Period of release, 22 sec

'2. RCIC steam line:.

Maximum release rate, 25 lb/sec of steam Period of release, 17 sec i

i. 3. RHCU piping:

Maximum release rate, 250 lb/sec of steam l Period of release, 22 sec i

~

Steam leakage into the Reactor Building could be exhausted through the ventilation exhaust systems operating at normal building pressures at a calculated rate of 63 lb/sec of steam. The SGTS operating at normal i i Reactor Building pre;,sures could exhaust about 5 lb/ set of steam.

3 Steam leakage in excess of these amounts would result in- Reactor j Building pressure increases above normal.

$ The Reactor Building is designed to relieve excessive internal l pressures so as to preserve main structural integrity, considering the l

rapid pressure reduction outside the building associated with i tornadoes. Refer to Appendix H.5. Reactor Building differential

[ pressures exceeding about 0.5 psi will be relieved through the Reactor i Building roof.

Steam leakage within the normal operating capability of the

ventilation exhaust systems would be ducted to the main building i exhaust vent. The ventilation exhaust from the principal Reactor '

, Building compartments housing the RCIC steam supply line and turbine, the HPCI steam supply line and turbine, and the RHCU are monitored by

. temperature elements. These elements provide temperature indication

and high temperature alarms in the main control room. The temperature i set points for these' alarms will alert the operator to potential steam

leakage conditions at leakage rates that are less than the normal j operating capability of the ventilation exhaust systems.

Steam leakage rates that exceed the capability of the normal operating ventilation exhaust systems could result in abnormal ventilation flow paths and abnormal Reactor Building exhaust locations. The design of

the Reactor Building would indicate that likely abnormal release y 5.3-6 Revision 9 - July 1988 en - , - - , , - , . c

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

PNPS-FSAR i locations 'would include the building roof and building access 7

locations.-

Steam leakage into a compartment within the operating capability of the ventilation exhaust systems ~ would be confined -within the normal

. exhaust paths, and therefore would limit the steam flooding principally to the compartment where the leakage originated. Thus the operability of safety related equipment, controls, and instrumentation located in other compartments would be maintained.

2 The ventilation exhaust temperature sensors will detect steam leakage

- from the RCIC steam line, the HPCI steam line, or the RHCU piping at

- leakage rates that are. below the normal operating capability of the ventilation exhaust from the compartments housing these hot, i

j

! ^ pressurized lines. Early detection of steam leakage at rates below  :

l the_ capability of the normal ventilation systems and subsequent J

!- isolation of leaks provide protection of safety related equipment within the Reactor Building. See Section 7.3.

I The : main stack provides an elevated release point for airborne activity during the postulated station loss of coolant and refueling

, accidents. Release of activity _ to the environs from the Secondary -

l Containment System is analyzed in detail in Section 14. Station Safety' Analysis. It'is concluded that the safety design bases are met.

5.3.5 Inspection and Testing

) The secondary containment leakage rate can be determined in the 4 following manner. The Reactor Building is isolated and the SGTS is started with one treatment train and its associated exhaust fan. The i

exhaust flow rate is controlled by the fan outlet damper control position as determined by flow rate measurements in the SGTS exhaust s

duct. The fan outlet damper positioner is used to control the exhaust j flow rate at 4,000 ft /3min.

a If the subatmospheric pressure as measured within the Reactor Building is equal to or exceeds 0.25 in of water (with neutral wind conditions at the site) the building safety design basis leaktightness with respect to inleakage is verified.

Tests of the ability of the various isolation initiation signals to

, automatically render the Reactor Building isolated, to trip the supply 4

and exhaust fans, and to start the SGTS can be conducted by simulating

the isolation signals.

Provisions are made for periodic tests of each filter unit. These 4

tests include determinations of differential pressure across each filter and of filter efficiency. Connections for testing, such as injection and %mpling, are located to provide adequate mixing of the injected fluid and representative sampling and monitoring, so that test results are indicative of performance. Each HEPA is tested with DOP (Di-or.tyl-phthalate) smoke. The charcoal filters can be tested for bypass with freon.

The electric heating coil in each filter train is tested to show that i the relative hunidity of an entering air stream is reduced.

5.3-7 Revision 9 - July 1988 w r .

PNPS-FSAR 5.3.6 Nuclear Safety Requirements for Plant Operation General The entries in this section represent the nuclear safety requirements for the SCS for each BHR operating state which represents an extension of the stationwide BHR system; analysis of Appendix G. The following referenced portions of the Safety Analysis Report provide important information justifying the entries in this section:

Reference Information Provided

1. Earlier parts of Section 5.3 Description of the SCS

2. Station Nuclear Safety Opera- Identifies conditions and tional Analysis Appendix G events for which the SCS is required

3. Jacobs, I.M. Guidelines for Describes methods used to Determining Safe Test Inter- establish allowable repair vals and Repair Times for times Engineered Safeguards. General i Electric Co., Atomic Power l

Equipment Department, APED - 5736, April 1969  :

Each detailed requirement in this section is referenced, if possible, to the most significant station condition originating the need for the -

requirement by identifying a matrix block on one of the Matrix 3 sheets of Table G.5-3. The matrix block references are given in parentheses beneath the detailed requirements in the " minimum required for action" section.

The matrix block references identify the BHR operating state, the event number, and the system number. For example, F40-91 identifies BWR operating state F (Matrix 3), event (row) No. 40, and system (column) No. 91.

System Action The SCS operates to limit the release of airborne radioactive materials to the environs.

Number Provided by D nign

1. One Reactor Building

2. One main stack

3. One RBICS, with two dampers in series provided throughout the isolation system to provide redundancy. The control system is designed such that all dampers fail closed on loss of dc power to the scienoids or on loss of instrument air to the valves.

5.3-8 Revision 12 - Jan 1991

t PNPS-FSAR I

l l

l 4. One SGTS consisting of two identical, parallel air filtration assemblies and two full capacity exhaust fans.

Mini =n= Reauired for Action BNR Operating States A B.C,0,E, and F:

-The Reactor Building I

(A40-90) (B40-90)

(C39-90) (039-90)

(E39-90) (F39-90)

The Main Stack (A40-105) (B40-105)

(C39-105) (D39-105)

(E39-105). (F39-105)

One damper at each isolation point (with associated controls)

(A40-102) (B40-102)

(C40-102) (40-102)

(E40-102) (F40-102)

One filtration assembly train and one exhaust fan (A40-91) (B40-91)

(C39-91) (D39-91)

(E39-91) (F39-91) 5.3.7 Current Technical Specifications The current limiting conditions for operation, surveillance requirements, and their bases are contained in the Technical Specifications referenced in Appendix B.

5.3-9 Revision 12 - Jan 1991

ljNFORMAT!ON~

, PNPS-FSAR , , , , ,

M i 9.4 GASEOUS RADWASTE SYSTEM L Use roso. .. . a j

. treferQ090 - (

l 9.4.1 Power Generation Objective 1

The Gaseous Radwaste System processes gaseous radioactive wastes from I the main condenser air ejectors, the startup mechanical vacuum pump, l the gland seal condensers, and other minor sources, and controls ]

their release to the atmosphere through the main stack in such a way that the operation and availability of the station is not limited.

4 9.4.2 Power Generation Design Basis

1. The Gaseous Radwaste System is designed so that gaseous and l particulate radwastes are processed and discharged such that l operation and availability of the station is not limited. i 1

2. The Gaseous Radwaste System is designed to minimize the possible f explosion hazard of the hydrogen and oxygen present. 1 11 9.4.3 Safety Design Basis

1. The Gaseous Radwaste System is designed to include equipment, instrumentation, and operating procedures such that the gaseous j radwastes can be discharged from the station at levels which are as low as reasonably achievable.

2. The Gaseous Radwaste System is designed to provide isolation on high offgas radioactivity level.

3. The Gaseous Radwaste System is designed to maintain its integrity l for all expected operating conditions by conservative process design.

9.4.4 Description 9.4.4.1 Air Ejector Offgas and Augmented Offgas System 9.4.4.1.1 General l

The Air E]ector and Augmented Offgas System shown on Figures 11.4-1 l and 9.4-1 includes the subsystems that process and/or dispose of the l gases from the main condenser air ejectors, the startup mechanical vacuum pump, and the gland seal condensers. All such gases from the I unit are routed to the main stack for dilution and elevated release I to the atmosphere. Discharges from the air e]ector, the charcoal vault, and the stack are continuously monitored by radiation monitors.

Gases routed to the main stack include air ejector and gland seal offgases, and gases from the Standby Gas Treatment System (SGTS).

Dilution air input to the stack is supplied by two full capacity fans located in the filter building at the base of the main stack. The stack is designed such that prompt mixing of all gas inlet streams occurs in the base to allow location of sample points as near the 9.4-1 Revision 2 - July 1983 1

l

PNPS-FSAR base .as possible. The stack drainage is routed to the liquid radwaste collection system.

As a design basis for the system, a noble gas input equivalent to an annual average offgas' release rate (based on 30~ min decay) of 100,000 microcuries/sec was used. Table 9.4-1 indicates the design basis neSle gas activity referenced to 30 min and the . noble gas activity

... . te r processing through the Augmented offgas System. Also shown on Table 9.4-1 are individual noble gas isotope activity reduction factors- and the overall activity reduction factor provided by the system. The' Augmented' Offgas- System receives the noncondensible gases. discharged from the main condenser. These gases and their volumetric flow rates are given on Table 9.4-2. The air inleakage

design basis for the Pilgrim Nuclear Power Station. Augmented Offgas System has been established at 7 fta/ min (at 130'F, 1 atm) .per condenser shell. Leakage from two condenser shells ft (corrected to 3/ min, the standard conditions) gives a total of 12.3 standard design basis air inleakage for the system.

Air inleakage at three operating boiling water reactors where condenser inleakage has a significant effect on offgas holdup time is given on Table 9.4-3.

9.4.4.1.2 System Function The Augmented Offgas System shown on Figure 9.4-1 uses a high temperature catalytic recombiner to recombine radiolytically

,. dissociated hydrogen and oxygen from the Air Ejector System.

Noncondensible radioactive offgas is continuously removed from the main condenser by the air ejector during plant operation. The air ejector offgas normally contains activation gases, principally N-16, 0-19, and N-13. The N-16 and 0-19 have short half lives and quickly decay. The 10 min half-life N-13 is present in small amounts which is further reduced by decay. The air ejector offgas also contains

the radioactive noble gas parents of biologically significant Sr-89, Sr-90, Ba-140, and Cs-137. The concentration of these noble gases

depends upon the amount of tramp uranium in the coolant and on the reactor fuel cladding surfaces (usually extremely small) and the

+

' number and size of fuel cladding leaks. After hydrogen / oxygen recombination and chilling to strip the condensible to reduce the volume, the remaining noncondensibles, principally kryptons, xenons, ,

and air, are delayed in a 30 min holdup system before reaching the adsorption bed. Radioactive particulate daughters of the noble gases are retained on the HEPA filters and on the charcoal. The charcoal adsorption bed, operating in a constant temperature vault, selectively adsorbs and delays the xenons and kryptons from the bulk carrier gas, principally air. This delay on the charcoal permits the Xe and Kr to decay in place. The offgas is discharged to the environs via the main stack. The activity of the gas leaving the Offgas- Treatment System is continuously monitored. This system results in a reduction of the offgas activity (curies) released by a factor of approximately 185 relative to a 30 min holdup system as shown on Table 9.4-1.

9.4-2 Revision 2 - July 1983

i PNPS-FSAR 4

The adsorption of noble gases on charcoal depends u; ion gas flow rate, holdup time, mass of charcoal and a gas unique coefficient known as the dynamic adsorption coefficient. The parametric inter relationships and governing equations are well proven from three years of operation of a

' - similar unit at KRB in Germany

' The design requirements for the equipment of the Offgas System are given on Table 9.4-4. The Augmented Offgas System is designated seismic Class II. This class includes those structures, equipment, and components which are important to reactor operation, but are not essential for preventing an accident which would endanger the .public health and safety, and are not essential for the mitigation of the consequences of these accidents. A Class II designated item shall not degrade , the integrity of any item'designatec Class I.

The front end components of the system are installed in the Turbine

Building. The charcoal adsorbers and associated auxiliary equipment are installed in the Retention Suilding whose access doors are at elevation 23 ft. As described in Section 2.4.4.3, station structures at elevation 23 ft are not subjected to flooding.

The system is not designed to be functional during or after a tornado.

The system is not essential for the prevention of accidents nor is it essential for the mitigation of the consequences of such accidents.

The Offgas System is provided with flow, temperature, and radiation instrumentation to ensure proper operation and control. Hydrogen analyzer instrumentation is also provided to ensure that hydrogen

. concentration is maintained below the flammable limit. Table 9.4-5 lists process system alarms and their location.

Offgas radiation monitoring is divided into two subsystems. One subsystem takes a continuous sample from the offgas line just after the air ej ectors . The other takes a contit.uous sample from the Offgas System just before discharge to the main station stack. The former subsystem is described in Section 7.12.2.

The subsystem monitoring the Offgas System upstream of the main station stack has two instrumentation channels. Each channel consists of a gamma-sensitive detector, a pulse preamplifier, a logarithmic radiation monitor with a power supply and a meter, and a strip chart recorder point. The monitors and the two-pen recorder are located in the control room. Each logarithmic radiation monitor is powered from a different bus of the 24V de system. The two gamma-sensitive scintillation detectors are mounted in two shielded sample chambers. The sample is drawn from the offgas line through the sample chamber by the sample pump. Each monitor has three upscale trips and a downscale trip. An upscale trip indicates high radiation. A downscale trip indicates instrument trouble. Any one trip will give an alarm in the control

. room. Any one upscale high radiation trip closes the charcoal bed filter bypass valve, if open, and opens the offgas line to the charcoal bed, if closed. Two upscale high-high radiation trips, 9.4-3 Revision 15 - June 1993

G PNPS-FSAR one upscale .- high-high' radiation trip and one downseale trip or . two downscale trips isolate . the . Offgas System outlet and drain valves (See Fi gure . - 11. 4 - 1 ) . Subsequent - operator action is ; required to reopen .the valves. This will ensure- positive control of releases to the environment.

The Offgas System radiation monitors have - monitoring characteristics sufficient . to provide accurate indication of radioactivity in the air ejector offgas, and provide the operator with sufficient information to

. monitor the performance of the Augmented Offgas System. Sufficient redundancy is.provided.to allow maintenance and testing of the Radiation Monitoring System. Each channel can be calibrated by analyzing a grab sample.

Figure 9.4 2 shows 'the location and arrangement of the front end components of the Augmented Offgas System in. the Turbine Building.

Figure 9.4 3 shows the arrangement of the charcoal adsorbers and l adsorber auxiliary equipment in a building located approximately 20 ft

-south of the Turbine Building between column . lines 3 and 8. The

. building is approximately 68 ft by 72 ft in plan and extends 20 ft above ,

and below grade.

'9.4.4.1.3 System Operation Noncondensible gas removed from the main condenser, including air in-leakage, is Oiluted with steam to less than 4 percent by volume hydrogen

. concentration in the steam jet compressors. See Figure 9.4-1. The diluted offgas is superheated and then passed through a catalytic .

recombiner to convert the hydrogen and oxygen into water. The offgas  :

effluent - from the recombiner, containing only traces of hydrogen, is ,

passed through a condenser, cooled by condensate , to remove the bulk moisture, and then to a.30 min holdup for the decay of the N-13, N-16, ,

0-19, krypton and xenon isotopes. Decay daughters and iodine are removed by condensation on the walls of the holdup pipe and further removal of the decay daughters is effected by filtration. The offgas is processed by a cooler condenser to remove additional moisture and iodine, a deentrainer and reheater to reduce the relative humidity, and a high efficiency filter prior to entering the charcoal adsorbers.

The charcoal adsorbers provide further delay of the xenon and krypton isotopes in the offgas. Two parallel trains of adsorbers are used to minimize back pressure. Heat is removed from the vault housing the adsorbers to maintain the charcoal beds at an approximate constant temperature of 77'F. The offgas effluent from the adsorbers is passed through another high efficiency filter prior to discharge to the offgas stack.

No -dilution air is added to the offgas stream during steady state operation. The air present during operation is from air inleakage into the main condenser which operates at sub-atmospheric pressure. Oil free'

~ air is bled into the system during startup of the system. Its flow rate is 56.7 lb/hr, which is stopped after the recombiner 9.4-4 Revision June 1993 -

PNPS-FSAR.  ;

comes up to temperature. Air is supplied during recombiner startup in.

~

order to prevent wetting of the recombiner catalyst and subsequent deterioration of the hydrogen recombiner performance.

In the event of failure 7of a nonredundant Augmented .0ffgas System component, provisions are made to bypass the Augmented Offgas System and operate the station using the installed 30 min Offgas Holdup System until maintenance of the Augmented Offgas System can be

. completed. ,

9.4.4.1.4 Safety Evaluation ,

~

The decay time prov%ed . by' the Augmented Offgas (A0G) System permits l significant -radioactive' decay of the activation gases and fission gases in the main condenser offgas prior to release. The design basis holdup is 22 days for xenon. isotopes and 29 hr for krypton isotopes. .

The function of the A0G system is to limit offsite releases to provide .

assurance that releases of radioactive gaseous effluents will be kept-As1 Low As Reasonably Achievable. Operation of the system with hold up .i times which are less- than the design values will reduce the 1 operational capability of the - system but -it does not result in a condition outside the analyzed bounds so long as the Technical Specification limits on release rate remain unchanged. Operation below- '

the solid line in Figure 9.4-5 will provide assurance that release rates remain -below Technical Specification limits and operational objectives. The ratio of Kr holdup time to Xe holdup time on Figure '

9.4-5 is.1:17. The Steam Jet Air Ejector monitor is located prior to

- the A0G charcoal beds. The solid daughter . products of the decay of

' the noble gases are removed by filtration and/or are- retained on the charcoal. Final filtration of the charcoal adsorber effluent precludes escape of charcoal fines. Particulate activity release is expected to be negligible, f

l I

r 9.4-4a Revision 13 - June 1991 ,

i

PNPS-FSAR Iodine input into the offgas system is small because of its retention

.in reactor water and condensate. Additional iodine removal is provided by steam condensation which occurs in the offgas condenser located downstream of the hydrogen recombiner. Minute quantities of iodine entering the charcoal adsorbers are further adsorbed.

The Augmented Offgas System radiation monitors, which are normally monitoring the release rate from the adsorber beds, can be selectively valved to monitor the release rates from the recombiner outlet, the HEPA filters' outlet, or the outlets of the first charcoal bed in each train, all of which provide diagnostic information on the performance of the charcoal bed Holdup System. The charcoal bed adsorber radiation monitor also automatically isolates the Offgas System in the event of high radiation levels in order to prevent treated gas of unacceptably high activity from discharging to the atmosphere.

Shielding is provided for offgas system equipment to maintain safe radiation exposure levels for plant personnel. The equipment is principally operated from the control room.

The charcoal adsorbers operate in a massive temperature controlled vault of 77'F so that upon system shutdown, radioactive gases on the adsorbers will be subject to the same holdup time as during normal operation, even in the presence of continued air flow. The adsorbers are maintained at a constant temperature by an air conditioning system which removes the decay heat generated in the adsorbers. Failure of the Air Conditioning System will cause an alarm in the control room.

In addition, a radiation monitor is provided to monitor the radiation level in the charcoal bed vault. High radiation will cause an alarm in the control room.

Th0 hydrogen concentration of the gases from the air ejector is

'"di ntai n ed below the flammable limit by maintaining adequate steam flow for dilution at all times. This steam flow rate is monitored and alarmed. The preheaters are steam heated rather than electrically heated in order to eliminate the presence of potential ignition sources and to limit the temperature of the gases in the event of cessation of gas flow. The recombiner temperatures are monitored and alarmed to indicate any deterioration of performance. A hydrogen analyzer downstream of the recombiners provides an additional check.

The Air Ejector Offgas System operates at a pressure of about 5 psig or less so the differential pressure which could cause leakage of radioactive gases is small. To minimize the possibility of leakage of radioactive gases, the system is welded wherever possible and bellows seal valve stems or equivalent are used.

Operational control is maintained by the use of radiation monitors to assure that the release rate is within the established limits.

Environmental monitoring is used to determine resultant dose rates and to relate these to the release rates as a check on station performance. Provision is also made for sampling and periodic analysis of the influent and effluent gases for purposes of 9.4-5 Revision 13 - June 1991

PNPS-FSAR determining their composition. This information is used in calibration of the monitors and in relating the release to environs dose. The operator is thus in full control of the system at all times.

9.4.4.1.5 Halfunction and failure Mode Analysis Table 9.4-6 contains a detailed malfunction and failure mode analysis indicating the consequences of failure of various components of the system and design precautions taken to prevent such failures.

9.4.4.1.6 Inspection and Testing In the Augmeated Offgas System calibration and maintenance of monitoring equipment' is performed on a routine basis. However, various temperatures, flow rates, and level signals are continuously monitored to detect for possible system malfunctions. Remote multiplexers transmit these signals, on demand, to a receiver in the main control room and on to the system computer. The computer calculates various parameters and sequences them for display on led readouts and on a multipoint trend recorder.

The particulate filters are tested after installation using a diocty1phthalate (D0P) smoke test equivalent. During operation, they will be periodically tested by laboratory analyres of inlet and outlet millipore filter samples ~.

Experience with boiling water reactors has shown that the calibration of the offgas and effluent monitors changes with isotopic content, isotopic leaks in the reactor, and the nature of the leaks. Because of this, the monitors are calibrated against grab samples periodically and at any time there appears to be a significant change in offgas release rate.

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9.4-6 Revision 13 - June 1991

PNPS-FSAR 9.4.4.2 Turbine Sealing and Mechanical Vacuum Pump Systems l 9.4.4.2.1 System Function The Gland Seal Holdup System collects and processes by delay the noncondensible exhaust from the main turbine gland seal condenser.

During startup operation the discharge of the condenser mechanical vacuum pump is routed through the Gland Seal Holdup System. The effluent of -the Gland Seal Holdup System is routed to the rain station stack where it is continuously monitored by the main stack Radiation Monitoring System before discharge to the environment. See Section 7.12.3.

During normal operation of the Gland Seal Holdup System, a 2,200 lb/hr saturated air-water vapor mixture containing trace amounts of hydrogen, oxygen, and radioactive gases is exhausted from the turbine generator glar.d seal condenser and enters the.16 in dia holdup line.

After being delayed for a' period of approximately 1.75 min, the effluent is routed to the main stack where it is mixed with the Augmented- Offgas System effluent and the dilution fans before release to the environment.dischargeofthestackl See Figure 11.4-1.

9.4.4.2.2 System Operation The Gland Seal Holdup System shares with the Augmented offgas System the main stack, dilution fans and the stack Radiation Monitoring System.

During normal operation, the amount of radioactive activation and fission gases associated with the Gland Seal Holdup System is extremely small. The radioactivity that is collected and processed by the Gland Seal Holdup System is proportional to the amount of main steam utilized in the main Turbine Sealing System. This amount of steam is less than 0.1 percent of the full power rated steam flow.

In addition to the small amount of radioactivity processed, there is j a correspondingly small amount of radiolytic hydrogen and oxygen '

which are well below the explosive limits.

The Gland Seal System is designed to provide a 1.75 min holdup delay time for the radioactive gaves before discharge to the main stack. i This design is consistent with maintaining discharges within  !

alloweble limits due to the extremely small amount of radioactivity  !

associated with this system.  ;

i During startup operations, the condenser mechanical vacuum pump is used to assist the steam jet air ejectors in achieving condenser vacuum. The discharge of the mechanical vacuum pump is routed through the Gland Seal Holdup System. The holdup normally provided 1 by the Gland Seal Holdup System is reduced.during startup due to l higher air throughput when the mechanical vacuum pump is operating.

Because the radioactive gases in the main condenser during startup are only a small fraction of the design evolution rate. the effect on radioactive effluents released to the environment is negligible, o

9.4-7 Revision 2 - July 1983 i

l

PNPS-FSAR The magnitude of the sources and resultant site boundary exposures resulting from station 'startups utilizing mechanical vacuum pump operation are difficult to quantify because the number, nature, and duration of the preceding shutdowns are difficult to estimate. An order of magnitude estimate of the annual average exposure from ten startups per yr was performed, assuming 4' hr of mechanical vacuum pump operation per startup, which indicated that maximum whole body exposures are less than approximately 0.05 mrem /yr. These estimated exposures. can be- reduced by minimizing the duration of mechanical vacuum pump operation.

9.4.4.2.3 Safety Evaluation -

The amount of- radioactivity associated with the Turbine Sealing System is negligible. The extremely low ' levels of radioactivity released from the Gland Seal Holdup System make direct radiation monitoring impractical; therefore, the total stack effluent is continuously monitored by the stack Radiation Monitoring System.

Excessive release of radioactivity from this system is not considered credible due to its passive design and the small amount of main steam utilized in the sealing process.

The Gland Seal Holdup System is a passive system operating at atmospheric pressure requiring no particular control or instrumentation. Monitoring of the Gland Seal Holdup System effluent is provided by the stack Radiation Monitoring System which monitors the combined effluents of the Gland Seal Holdup System and' Augmented offgas Fystem.

9.4.4.2.4 Inspection and Testing

'he Turbine Sealing System is continously operated during station operation and does not require specific testing to ensure

' operability.

9.4.4.3 Miscellaneous Gaseous Effluents 9.4.4.3.1 Low Release Potential Effluents Miscellaneous gaseous effluents are categorized into two classes, those from areas having a negligible or low potential for the release of airborne radioactivity, and those from areas likely to experience radioactive contamination. Following is a list of station areas which fall into these catagories. These areas are exhausted directly l to the environment.

1. Diesel Generator Building

2. Administration Building

3. Machine Shop

4. Battery Room and Lube Oil Compartments J

9.4-8 Revision 2 - July 1983

PNPS-FSAR l

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5. Recirculation Pump HG Set Area

6. Turbine Building Operating Floor and Switchgear Area The ventilation air from . the f.irst five areas listed above has a negligible potential for the release of radioactive effluents. The  !

Turbine Building operating floor including the reactor -feedwater pump area are considered to have a low potential for release. Any release l from the. Turbine Building basement area or the Turbine Building ground floor to the Turbine Building ' operating floor or adjacent areas above elevation 51 ft is, precluded since the Turbine Building j basement and ground floor are maintained at a slight negative

pressure relative to the Turbine Building operating floor.

4 The airborne radiation concentration levels at elevation 51 ft in the Turbine Building are routinely nonitored by means of a constant air monitor with local alarm and recorder. Airborne activity levels in

those areas of the station having a direct release path to the 4

environs not monitored by a process radiation monitoring system will

, under normal operating conditions be within those levels allowtd for in Appendix B, Table I, of 10CFR20, revised as of January 1,1970.

Assuming release from the Turbine Building operating floor at concentrations up to 3 X 10-9 microcuries/cm3 the resulting concentrations at the site boundary would be less than 10-3 to 10-2 i

of MPC, on an unidentified basis.

T[1e expected airborne activity on the Turbine Building operating floor will normally be below the values assumed above and the i releases from the Turbine Building operating floor and the reactor '

feedwater pump area are expected to be insignificant relative to the releases from the main stack and the Reactor Building exhaust vent.

. 1 9.4.4.3.2 Potentially Contaminated Effluents , ,

Gaseous effluents from areas of potential radioactive contamination i

are monitored and discharged to the environment through either the -

main stack or the Reactor Building exhaust vent. See Figures 10.9-4,

~

i 10.9-5, 11.4-1, and Section 10.9. The station ventilation systems (

are designed to combine the ventilation air flow from these areas and i exhaust that air past process radiation monitoring equipment. The 1 operation of the process radiation monitoring equipment is described {

j, in Sections 7.12.3 and 7.12.5.

Miscel,1laneous sources of potential low level radioactive airborne  :

l

, contaminants in the station which could be released to the environment are:

a. Primary Conthinment Venting l

Primary system leakage inside the primary containment could i 4

occur as a result of recirculation pump seal leakage , valve flange leakage, and valve stem packing leakage. The latter two are also sources of potential leakage outside the primary containment. The magnitudes of these leaks are 9.4-9 Revision 2 - July 1933 h

I j

!- PNPS-FSAR i

minimized. to the' extent possible by. regular periodic I

inspections and station maintenance activities. 1 the site An analysisresulting ~ was performed ifrom primary Lto estimatecontainment ., boundary purging l

exposures

assuming a 5-gal / min unidentified steam leak for a period of

' time sufficient to reach equilibrium concentrations in steam l

. equivalent to a 25,000 microcurie /sec offgas rate after

- 30 min decay _ was used. The- resultant whole body exposure i

per purge is estimated 'to be 'less ~ than 0.001 stem assuming -

that purging commences after the reactor has been brought to hot standby. During station- operation, the drywell atmosphere is sampled for activ,ity level to ensure that i-releases from this source will be minimal. ..

, b. Steam Leakage Outside the Primary Containment The site bouadary exposures resulting from steam leakage outside the , primary containment have been estimated based i upon- releases equivalent to a continuous steam leak of 7 gal / min of saturated liquid from the station ventilation i

exhaust. This - has been selected based on experience at ,

operating plants. The release to the environment may occur i

l i

I. from the Turbine Building roof vent or the reactor building >

exhaust vent. Upper estimates of~the magnitude of the whole l

body exposure resulting from noble gas releases from the n steam range from 0.1.to 0.4 millirem /yr.

t Assuming a leak rate of 7 gal / min and a coolant

concentration concentration consistent with an offgas 4

release rate of 25,000 microcuries/sec as measured at 30 min

! decay and a condensat' ion plateout factor of 2 results in an -

environmental release rate of 0.04 microcuries/sec of I-131 with corresponding releases of I-132 to I-135. The value of  !

0.04 microcuries/sec release rate for I-131 can be compared j to measurements which have been made on operating BWRs which .

have shown release 10ges ra from the building ventilation 1

systems of 2 x - microcuries/sec to 4 x 10-2 microcuries/sec. The rate of release predicted results in a l site boundary exposure rate of 0.6 millirem /yr.

t+

i c. Tank Vents and Sumps 4

,. Vents from liquid waste storage tanks, aerated resin

]- regeneration tanks, open equipment, and floor drain sumps provide very little potential for contamination release, as

the quantities of radioactive noble gases present in these

! liquids are generetly negligible.

i i Particulate activity in the air space above stored liquid radwaste solutions is related to the gas liquid partition l coefficient at the air water interface. The magnitude of this coefficient coupled with the filtration of a majority of the vents through high efficiency particulate (HEPA) 9.4-10 Revision 7 - July 1987 i

e - -__ -__________.___ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _- _ _ _ _ . _ _ _ _ _ _ ---- *--

l PNPS-FSAR 1

i filters minimizes the potential for particulate releases from liquid waste storage tanks and open cumps. .The collection of ionic halogens on station demineralizer resins

- creates an additional potential source of noble gases through the decay of halogens to their noble gas daughters.

However, these gases are not released promptly to the environment. The operation of the liquid radwaste system minimizes the release of gaseous daughter fission products by allowing the system components to act as gaseous delay tanks to effect the decay of the significant noble gas daughters. Only the occasional necessary air scrubbing and air sparging of certain radwaste tankage and normal tankage filling provides potential release mechanisms. The site boundary dose contribution from these sources is expected to be negligible.

d. Hood Vents Radiochemical hood vents provide a potential miscellaneous source of release of airborne activity from the station.

However, the sampling frequencies and volumes result in releases which are small fractions of the releases from other miscellaneous sources from the station. Further, the ,

HEPA filters installed in the exhaust ducting from the radiochemical hoods act to ensure that no particulate radioactivity is released.

1

e. HPCIS Testing ,

1 The site boundary exposure due to testing of the High Pressure Coolant Injection System (HPCIS) for an assumed 30 hr/yr has been evaluated. The HPCIS turbine uses primary system steam for motive force of which 500 lb/yr is used as l HPCIS turbine gland sealing steam and is condensed in the HPCIS gland seal condenser. The associated noncondensibles including trace amounts of noble gases are released during test operation .to the environment through the reactor building exhaust vent or the Standby Gas Treatment System (SGTS) which discharges to the main stack. The resultant site boundary whole body exposure is negligible less than approximately 1 percent of that expected due to primary system leaks outside the primary containment.

9.4.5 Estimates of Radioactive Gaseous Releases During Normal Operation Estimates of radioactive gaseous releases and resultant doses during normal operation are given in the Pilgrim Station Unit 1 Appendix I Evaluation, dated April 1977.

9.4-11 Revision 2 - July 1983

a PNPS-FSAR E

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L._We rGSh h; 7.9 A 10.9 HEATING, VENTILATION, AND AIR CONDITIONING SYSTEMS reference >

10.9.1 Power Generation Objective '

The power generation objective of the Heating, Ventilation, and Air Conditioning (HVAC) systems is to control the station air temperatures .

and the flow of airborne radioactive contaminants to ensure the operability of station equipment, and the accessibility and habitability of station buildings and compartments.

10.9.2 Power Generation Design Basis The HVAC systems:

1. Provide temperature and humidity control and air movement for personnel comfort and optimum equipment performance

2. Provide a sufficient filtered fresh air supply for personnel

3. Provide for air movement from lesser to progressively greater areas of radioactive contamination potential prior to final exhaust

4. Minimize the possibility of exhaust air recirculation into the air intake 10.9.3 Description 10.9.3.1 General The HVAC systems provide individual air supplies to main areas of the station. Normal airflow is routed from lesser to progressively greater areas of radioactive contamination potential prior to final i exhaust.

Most of the heating and ventilating systems utilize 100 percent outside air with 'no recirculation. All inlet air can be temperature controlled with heating coils.

Air from areas containing potential sources of radioactive contamination such as the Reactor Building, Radwaste Building basement, Turbine Building basement, and Radwaste Heating and Ventilation Building are discharged through .the building exhaust vent. Air from other areas is discharged at building roof levels.

The Equiprent Area Cooling System (EACS) is described in Section 10.18. Main control room heating, ventilation, and environment control are described in Section 10.17.

l 10.9.3.2 Station Heatin; System The station is heated during plant operation by a forced circulatior, hot water system. See Figures 10.9-1 and 10.9-2. The syster 1

10.9-1 l

l

PNPS-FSAR l

, ' consists of two package boilers, five hot water circulating pumps, two fuel oil transfer pumps, two 25,000 gal fuel storage tanks, one compression tank and associated piping, valves, . combustion controls, and instrumentation.

Heating system water is heated by two oil-fired package boilers. The boilers are located in the Reactor Building apxiliary _ bay. The hot water boilers are each rated at 16.7 X 10' Etu/hr at 125 psig,  :

utilizing No.-2 diesel fuel oil. l l i

The station heating system utilizes a low temperature (210*F max.),

two pipe, forced circulation hot water system. System water J tecoerature is automatically contro11ed based on outside air temperature. Water at temperatures adequate to maintain space temperatures is provided to unit heaters located throughout the station. -

t Three main system circulating pumps, each sized for one-half of the total design flow rate, provide continuous flow throughout the system. Normally, two run with one standby. Two booster pumps sized for full Turbine Building hot water coil capacity will provide

' additional capacity during low temperature periods. One pump runs with one pump a standby. The main heatinr piping headers are cross connected to allow isolation of either boiler when their use is not required. The auxiliary boilers and circulating pumps will be accessible for maintenance.

Heating system water temperature control is provided by a three-way mixing valve located in the hot water supply line on the suction side of the heating system circulating pumps. The actuating device for the three-way valve is an outdoor air sensing element controlling the position of the valve mixing unit to blend boiler supply and return water. This varies system water temperature inversely with outdoor temperature to reduce fluctuations in space temperatures as demand varies. During mild weather periods the boilers have a reduced flow; however, the heating input is reduced proportionately. The auxiliary boiler water temperature is controlled by a temperature insertion element controlling a modulating oil burner metering assembly.

Heating and ventilating equipment hot water coils have constant flow during pump operation. Leaving air temperature is controlled by an air temperature sensor modulating face and bypass air dampers.

Air conditioning fan unit hot water coils leaving air temperatures is controlled by a three-way valve located in the return line. Unit heaters have constant flow through their coils at all times during pump operation. Space temperatures are controlled by a space thermostat controlling the unit heater fan motor.

10.9-2 Revision 10 - July 1989

l PNPS-FSAR Minter design temperatures for the system are given on Table 10.9-1 and summer design temperatures are given on Table 10.9-2.

10.9.3.3 Reactor Building Heating and Ventilation System The Reactor Building. is divided into three major ventilation zones.

One zone encloses the spaces above the operating -(refueling) floor.

The second zone encloses the recirculation pump motor generator sets i

i and the third zone encloses the remainder of the Reactor Building.

Each zone is served by its own air supply and exhaust system in order i

l-to maintain the independence of the zones. See Figure 10.9-3. The i

systems for the first and third zones employ once-through ventilation without recirculation. The design basis summer maximum temperatures i

are area specific (See Table 10.9-2). .

{

i Thg/

ft min of filtered and tempered outside air which enters the React Building through louvers in the east wall. Two supply fans, each rated at design capacity, are located in the Reactor Building in a fan room. The fans are manually started. For normal operation one fan i

runs while the other fan is on standby. However, both fans can be during operated in parallel to provide additional ventilation refueling operations. If the operating fan fails during normal i i

operations, the standby fan starts automatically and an alarm is i received in the main control room. Air is exhausted from the operating floor through ducts located in the roof truss area and the south wall area adjacent to the floor. Additional exhaust ducts are located above the water level in the fuel pool, steam dryer, and separator pool, and in the reactor cavity. Two operating floor exhaust fans, each rated at design capacity, are located in the main fan room outside the Reactor Building. These fans discharge into the j main exhaust plenum at the base of t%

• tilding vent. The building vent is square in cross section exte w from . he t top of the main i exhaust plenum to the discharge point at elevation 182 ft asi.

The zone enclosing the recirculation pump motor generator sets is l

normally supplied with 50,000 ft 3/ min of filtered outside air or i

a recirculated air. Outside air enters through louvers in the west l wall. Two supply fans, each rated at decign capacity, are manually

] started. For normal operation one fan runs with the other as i standby. Both fans can be operated in parallel to provide additional Ventilation if required. If the operating fan fails, the standby fan l

starts automatically, and an alarm _ is received in the main control room. Two exhaust fans, each rated at design capacity, recirculate or l'

exhaust zone air. Exhaust air is discharged through louvers in the north wall. Temperature control is provided by an air temperature sensor modulating the supply, exhaust, and recirculation dampers.

Supplementary cooling for summer conditions is provided by unit

coolers supplied by the Reactor Building Closed Cooling Water (RBCCH)

System. iour unit coolers, each rated at half design capacity, are installed. Unit heaters are installed to provide zone heating during j winter shutdown conditions. The heaters are supplied by the station i heating system. l I

i 10.9-3 Revision 11 - July 1990 4

PNPS-FSAR der of the Reactor Building is supplied with a total of The remaig/

60,000 ft min of filtered and tempered outside air which enters the I, Reactor Building through louvers in the east wall into two fan rooms.

Each fan room contains two supply fans rated at design capacity. Each fan is capable of supplying 30,000- ft3 / min of-air to. various reactor building areas. The fans are manually started, and for normal '

operation, one fan in each fan room runs while the other fan is at standby. If the operating fan fails during normal operations, the standby fan starts automatically, and an alarm is received in the main control room.

The air supplied to the Reactor Building from the two operating supply fans is routed through the building to areas of progressively higher contamination potential. Air exhausted from areas of higher contamination potential (contaminated area exhaust) is routed independently of the exhaust from areas of expected lesser contamination (clean area exhaust).

Two contaminated area exhaust fans, each rated at design capacity, are located in the Reactor Building. The fans discharge to the main exhaust plenum at the base of the building vent. An addit.ional smaller exhaust fan located in the Reactor Building, exhausts only from the control rod drive maintenance shop, and discharges to the main exhaust plenum. Constant volume control is maintained by inlet vanes which are automatically positioned.

All the Reactor Building supply fans are electrically interlocked with )

their corresponding exhaust fans and run only when their associated

exhaust i.ns are op1 rating.

j 10.9.3.4 Turbine Building Heating and Ventilation System The Turbine Building air flow diagram is shown on Figure 10.9-4.

l 10.9.3.4.1 General l The Turbine Building Ventilating System supplies filtered air to all j areas of the Turbine Building and is routed to areas of progressively

greater radioactive contamination potential prior to final exhaust.

The ventilation system supplies filtered and tempered outdoor air to a the operating floor and all other areas below the operating floor. The j main condenser area is maintained at a slightly negative pressure to i prevent the spread of radioactive contaminants to the adjacent operating areas.

l l The exhaust air from the Turbine Building operating floor and switchgear areas is discharged to the atmosphere through exhaust fans located on the Turbine Building roof. The exhaust air from the reactor feedwater pump area will be discharged to atmosphere through exhaust fans located on roof above. The exhaust air from the condenser compartment and other adjacent potentially contaminated i

areas will be discharged through the building vent. The exhaust air I from the battery rooms and lube oil compartments will be discharged 10.9-4 Revision 11 - July 1990

1 INPS-FSAR ')

l by indepsnd:nt axh ust fcns leccted en ths Turbin2 Building cparating l floor through ductwork to the Turbine Building roof.

The condenser and condensate pump compartments depend on fan coil cooling units to supplement the main ventilation system when outside air ,

temperatures are above 60F.

10.9.3.4'.2 Main Turbine Building Ventilation Supply Fans 1

There are three Main Turbine Building ventilation supply fans, each of .

half capacity. Under normal operating conditions two fans are running, with one at standby. The fans are shut down in the event of loss of offsite power. During normal operations, any two fans are started. If an operating fan fails, a flow switch will sense a reduction of pressure and annunciate in the main control room. The standby fan is then started manually. If the fan discharge air temperature drops below 40F, a temperature switch' will stop the fans and annunciate in the ., main control room.

Heating coils are provided to heat the outside supply air when necessary. The supply air temperature is controlled by modulating the amount of air flow over the heating coils with face and bypass dampers.

Hot water flow to the heating coils is constant. The supply air volume is constant, with no recirculation. Outside air inlet dampers are interlocked with each supply fan and close if all fans are off. Outside air dampers -are heated with resistance cable to prevent freezing.

Constant volume is maintained by inlet vanes which are positioned automatically.

10.9.3.4.3 Turbine Building South Wall Ventilation Louvers (Operating Floor Level)

There are 3 louvered openings in the south wall of the Turbine Operating Floor. Each is 5' x 6', and they were originally designed to provide additional cooling for the operating floor during unusually warm weather. The unit on the west side has been permanently blocked off and the other two dr.mpers are normally closed but can be manually opened if needed.

10.9.3.4.4 Turbine Building Basement Exhaust Fans (Condenser Compartment)

There are three half capacity Turbine Building basement exhaust fans which discharge to the building vent. The fans are shut down in the event of loss of offsite power. The fans are started manually. If an operating fan fails, a flow switch will sense a reduction of pressure and annunciate in the main control room. The operator will then start the standby fan manually. Constant volume control is maintained by inlet vanes which are automatically positioned. The air volume exhausted by the fans is sufficient to maintain a negative pressure relative to all adjacent areas.

10.9-5 Revision 16 - June 1994 4

PNPS-FSAR 10.9.3.4.5 Turbine Building Operating Floor Exhaust Fans 4

Nine direct drive fans (two are spares) are located on the roof, and seven run during normal operation. The fans are shut down in the event of loss of offsite power. The fans are manually started. Fan motor malfunction is annunciated in the main control room.  :

10,9.3.4.6 Lube 011 and Battery Room Exhaust Fans Two full capacity fans are installed which exhaust to the Turbine Building roof. The fans are shut down in the event of loss of offsitea power. The fans are manually started. If the operating fan fails, flow switch will sense a loss of pressure and automatically start the standby fan and annunciate in the main control room. In order to detect fire damper failures, additional flow switches are provided to sound an alarm in the control room upon loss of flow in each Battery Room Exhaust System. Constant volume control is automatically maintained by the inlet vanes.

10.9.3.4.7 Condensate Pump Room Unit Coolers Two full capacity fan unit coolers are installed.The The fans units are shut are manually down in the event of loss of offsite power.

started. If the operating fan fails, a flow switch senses reduction of pressure, automatically starts the standby unit, and annunciates in the main control room. A temperature element (with local transmitter) is located in the discharge of each unit to give temperature indication in the main control room. Cooling water supply is from the Turbine Building Closed Cooling Water (TBCCW) System.

10.9.3.4.8 Condenser Compartment Unit Coolers Six unit coolers (two are standby) are installed in the condenser compartment and are supplied with cooling water from the TBCCW System.

The fans are started manually. ,High compartment temperature will alarm in the main control room and a standby unit is manually started.

The leaving air temperature is adjusted by regulating the flow of cooling water to each unit cooler.

10.9.3.4.9 off-gas Retention Building The off-gas Retention Building is supplied 5400 CFM of air from the Turbine Building HVAC System and 100 CFM of air from process air systems. This is all exhausted back to the Turbine Building HVAC System (refer to Figure 10.9-5).

Off-ras Retention Buildine Unit Coolers Three 2 horsepower unit coolers handle air received by the Retention Building. One cooler provides 3060 CFM to the operating floor and control room and utilize the TBCCW System as a heat sink. The other two (one is standby) provide 2340 CFM to the charcoal vault (refer to Figure 10.9-5).

10.9-6 Revision 9 - July 1988

PNPS-FSAR Off-aas Retention Build'ina Exhaust Fans Two 5500 CFM exhaust fans (one is standby) draw air from the charcoal vault filters, equipment room, operating floor, process vents and

drains, and other building sources to the Turbine Building HVAC System (refer to Figure 10.9-5). ,

10.9.3.4.10 Off-gas Retention Room Unit Coolers .

Each Off-gas Recombiner Room contains a 5000 CFM unit cooler which d

circulates and cools room air. These units utilize the TBCCH System l as their heat sink.

10.9.3.5 Radwaste Building Heating and Ventilating System Radwaste area air flow diagram is shown on Figure 10.9-5.

10.9.3.5.1 General The Radwaste Building -Heating and Ventilating System maintains required space temperatures, provides adequate ventilation to remove heat rejected from operating equipment compartments, and provides i adequate supply and exhaust to maintain the direction of air flow from i lesser to increasingly greater areas of potential radioactive i

i-i i

4 l l t 4 l

1 l l i

l 10.9-6a Revision 9 - July 1988

_ _ _ . _ _ _ _ _ ___-- a_ _ . - - - _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ _ _ _ _ . _ _ . _ _ . _ _ _ _ _ - _ - _ _ . _ . _ _ . _ _ _ _

PNPS-FSAR contamination. Exhaust hoods with negative pressure are provided at locations where, under normal operation, contaminants could escape to the surrounding areas. A pipe manifold vent system is provided for expected contaminated tanks and equipment. Filtered and tempered air is supplied to all personnel occupancy areas, the monitor tank compartment, and treated waste holdup tank areas in sufficient -

quantity to maintain space temperatures and ventilation. The air is supplied through ductwork by two full capacity air handling units, (one is normally running and one is a standby).

Exhaust air from the contaminated equipment tank cells, the vent pipe manifold, and the baling machine area, and- other ventilation system exhausts are routed ~ through the two banks of exhaust air filter assemblies and discharged to the building vent.

10.9.3.5.2 Radwaste Heating and Ventilating Units Each of the two heating and ventilating units is full capacity. The fans are shut down in the event of loss of offsite power. The fans are manually started. If the normal operating fan fails, a flow switch senses a reduction of pressure and automatically starts the standby fan and annunciates in the main control room. The fan discharge air temperature is controlled by a thermostat through a limit sensor and controller which modulates the heating coil face and bypass air damper. The heating coil is supplied by the station hot water heating system. The outside air damper is heated with a resistance cable to prevent freezing.

10.9.3.5.3 Radwas ! Exhaust Fans and filters The two exhaust fans are each full capacity. The fans are shut down in the event of loss of offsite power. The fans are manually started. If the normal operating fan fails, a flow switch senses loss of pressure, automatically starts .the standby fan, and annunciates in the main control room. Constant volume is maintained by inlet vanes automatically positioned. Two half capacity filter assemblies are installed. Each filter assembly can be isolated manually. Prefilter and HEPA filter differential pressures are indicated outside the filter assembly ccapartment.-

10.9.3.6 Access Control Area Air Conditioning 10.9.3.6.1 General The access control area air conditioning system maintains ventilation and constant temperature and humidity in the access control area. See Figure 10.9-6. The equipment, ductwork, and controls are completely

- independent from other station HVAC systems. The system independence will insure uninterrupted operation during normal and shutdown modes.

The access control area is served by full capacity redundant units including dual duct (hot-cold) air conditioning units, reciprocating condensing units, and recirculation and exhaust fans.

10.9-7 Revision 11 - July 1990

PNPS-FSAR f

Conditioned air is ' distributed through ductwork to the mixing boxes

! and diffusers located in the various zones. The zones independent - )

! from each other are:

I Corridor Instrument Repair .

l Chemical Laboratory and Counting Rooms

! Frisking Area Decon Shower l- Dressing Area i

j H.P. Office i

Chemical Lab Expansion Undressing Area Air is recirculated from the instrument repair room only. All the other rooms will be kept at a positive pressure with respect to the chemistry laboratory, thus allowing air to infiltrate from the lobby in the upper level into the chemistry laboratory. The exhaust from the laboratory fume hoods is exhausted by a booster fan through the radwaste filtering system prior to release from the building vent. )

10.9.3.6.2 Access Control ArJa Air Conditioning Units Two full capacity units are installed. The units are shut down in the event of loss of offsite power. The fans are manually started.

If the normal operating fan fails, a flow switch will sense a ' loss of pressure, automatically start the standby fan, and alarm in the main control room. One full capacity air conditioning unit is provided for the H.P. Office area'on El. 37'-0". The unit is shut down in the event of loss of offsite power. The unit is manually started and stopped. Space temperatures will be controlled by electric duct heaters. Each room is provided with its own thermostat for individual control.

10.9.3.6.3 Access Control Area Recirculation and Exhaust Fans Two full canacity recirculation and two full capacity exhaust fans are installed The fans are manually started and are shut down in the event of loss of offsite power. The exhaust fans discharge to the building vent. The recirculation fans either recirculate the air or discharp to the environment at roof level. If the normal operating fan fails, a flow switch senses loss of pressure, automatically starts the standby fan, and alarms in the main control room. Portions of the air exhausted from the counting room and H.P. l

}

10.9-8 Revision 11 - July 1990

PNPS-FSAR 4

counting worke space is discharged into hot ~ machine shop on El.

23'-0". The remainder of.the air is returned to the air conditioning unit, mixed with outside air and redistributed to the space. The exhaust fan will shut down on loss of offsite power..

10.9.3.7 Intake Structure .

10.9.3.7.1 General The intake structure is comprised of five heating and ventilating areas: one area containing the salt service water pumps; two areas containing the condenser circulating water pumps; an area containing the fire pumps; and an area containing the hypochlorite storage tank.

See Figure 10.9-6.

Exhaust fans are provided for the ventilation of each area. The salt water service pump area has redundant fan units while the other areas l

are served by single fans. Unit heaters are provided for heating.

Air is supplied to the various areas from outside louvers. Outside

air dampers are heated with sresistance cable to prevent freezing of controls.

i 10.9.3.7.2 Service Water Pump Exhaust Fans l

Two full capacity fans are installed and either fan will provide the i required air flow for the pump rooms. The fans may be powered by the diesel generators. One fan is manually started, and should the operating fan fail, a flow switch senses loss of pressure, 1 automatically starts the standby far and annunciates in the main control room. Constant volume controi is maintained by automatically i positioned inlet vanes.

10.9.3.7.3 Other Area Exhaust Fans l

One full capacity fan is installed for each of the remaining intake i structure areas. The fans are manually started and are shut down in

the event of loss' of offsite power.

4 10.9.3.8 Harehouse and Machine Shop I

10.9.3.8.1 General Filtered and temperature controlled air is supplied to the tool room,

machine shop, and warehouse. See Figure 10.9-6. The air is supplied

' through ductwork by one air handling unit. Approximately 20 percent

- of the total supply air is exhausted through the exhaust hood over the decontamination trough, then into the radwaste air filtering unit.

The remainder is exhausted to atmosphere through louvers at roof level.

The equipment, ductwork, and controls are completely independent from t other HVAC Systems except for the decontamination exhaust to the radwsste area.

10.9-9 Revision 11 - July 1990

m PNPS-FSAR 10.9.3.8.2 Heating and Ventilating Unit One full capacity unit is installed. The fan unit is manually started )

and is' shut _ down in :the event of loss of offsite power. Fan failure is annunciated in the main control room. - The supply air temperature is controlled by modulating face and bypass air -dampers. Hot water flow to heating coils is supplied by. the station heating system.  ;

Supply air flow is straight through with no recirculation. The

- outside air inlet dampers are interlocked with the supply fan and closed when the fan - stops. Outside air dampers' are heated with resistance cable to prevent freezing of controls.

10.9.3.8.3 Exhaust Fans Full capacity fans eihaust the warehouse and machine shop areas and

- the machine shop decontamination trough. The fans are started manually and shut down in the event of loss of offsite power. If the

area fan fails, a flow switch will sense a reduction of pressure and annunciate in the main control room.

1

! 10.9.3.9 Diesel Generator Building Heating and Ventilation 4

l Each standby diesel generator room is heated by hot water unit heaters j supplied by the Station Heating System. See Figure 10.9-3. Heating units are automatically shut down after the diesel generators start.

! Outdoor air is supplied to each diesel generator room through separate

! air intake plenums. Dampers control the air flow from the plenums I

! when a diesel generator is operating as follows:

i

!' 1. Outside Air Damper - This damper is associated with the air

- flow path which removes heat from the Diesel Jacket Water Cooling System during operation. The Diesel Generator's radiator fan draws air from a common intake plenum (which also serves the Ventilation Damper described below) through the

Outside Air Damper which is fully opened during Diesel Generator . operation. The spent cooling air exhausts from the building through the Exhaust Damper (also fully open) &

i exhaust louvers located near the roof. Both the Outside Air j and Exhaust Dampers are fully closed during shutdown. Neither

damper has a manual override and both fail open upon loss of j motive power or control signal to assure that cooling air is l

available.

2. Ventilation Air Damper - This damper is associated with the i

air flow path that provides general building ventilation as well as supplying the Diesel Generator engine with combustion "

! air. It is also sometimes referred to as the Combustion Air Damper. In shutdown mode, the Ventilation Air Damper is shut. Upon initiation of the engine start signal, this damper i fully opens to provide greater air throughout to support the engine vacuum. The Ventilation Damper has no manual override ,

l J and fails open upon loss of motive power or control signal in order to assure there i s always adequate combustion air )

available.

l-10.9-10 Revision 11 - July 1990

PNPS-FSAR Engine - freeze protection 'is provided by the Jacket Nater Cooling System' which maintains the engine coolant temperature at a constant.

value during' shutdown mode. Moreover, the engine coolant contains antifreeze in the event that coolant cannot be heated during shutdown mode.

Ventilation system ducts, dampers, fan,- and controls are Class I '

design.

10.9.3.10 Administration Building Air Conditioning The Administration Building is supplied by a hot and cold duct air-conditioning unit with air returned or exhausted by a recirculation fan. Conditioned air is distributed through ductwork to zone mixing boxes and ceiling diffusers.

During the winter, hot water baseboard convectors located along the outside walls help maintain the required temperature inside the building. Zone temperatures are controlled by a thermostat in each zone. The leaving air temperature of the heating and cooling coils is preset with conditioned air flowing to zone mixing boxes. Zone thermostats control metering valves in each mixing box to control air temperature. Base board convectors are controlled independently by a thermostatic valve in the return line to each zone.

The Administration Building Air Conditioning uses the TBCCW System as a heat sink.

10.9.3.11 New Administration and Service Building HVAC The New Administration and Service Building HVAC System is independent  !

to those of the process buildings.  ;

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l 10.9-11 Revision 11 - July 1990 i ,

I l

PNPS-FSAR TABLE 10.9-1 1 DESIGN TEMPERATURES (WINTER)

Outdoor: 10*F Dry Bulb Indoor: Minimum Turbine Building 60*F Reactor Building 60*F Access Control Area 75* F80 50%

relative humidity Administration Building 75* FBD 50%

relative humidity Radwaste Area 65*F Diesel Generator Building 60*F Intake Structure 60*F Machine Shop &

Harehouse Area 65'F Fire Hater Storage Tanks 45'F Demin. Water & Condensate Storage Tanks 45'F l

l l

1 of 1

PNPS-FSAR TABLE 10.9-2 DESIGN TEMPERATURES (SUMER)

Outdoor: 88'F Dry Bulb 74*F Het Bulb .

Indoor: Maximum Turbine Building 105'F - 120'F Reactor Building See Note (1 Below) l Control Room Area 78'FDB 50%

relative humidity l

i Access Control Area 78'FDB 50%

relative humidity t

1. Administration Building- 78'FDB 50%

3 relative humidity t

Radwaste Area 100*F Diesel Generator Building 105'F (Max)* - 95'F (5ft

- above Floor)* l i

Intake Structure 105'F Machine Shop and Harehouse Area 105'F Cable Spreading Room -102*F to +76*F l

NOTE 1: The following data represent the Reactor Building maximum

! summer design temperatures by specific location:

Max. Room Av. Room Temp (*F) l l

Location Temo. (*F) 5 Feet Above Floor Level i l

Refueling Floor 105 95 i 95

. General Floor Area 105 Nain Steam Pipe Tunnel 120 110 RHR/ Core Spray Pump Area 115* 100*

CRD Pump Area 115 100 RCIC Pump Area 115* 100*

HPCI Pump Area 115* 100*

100 Cleanup Regn. & Nonregn. 115 Heat Exchanger

• These temperatures apply only to conditions when the equipment in the area is operating. Under normal plant operation, temperature in these areas will be lower than indicated above.

1 of 1 Revision 11 - July 1990  ;

- , -, + -'

- ~ ~ ---- _

~~

~ ~c

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PILGRIM NUCLEAR POWER STATION l SEMIANNUAL REPORT

'/ I I

]

EtiVIRONf1 ENTAL RADIATION MONITORING PROGRAM<l July 1-December 8,1972 l

~

C Prepared for The Boston Edison Company 800 Eoylston Street Boston, Massachusetts 02199 Prepared by l l

25; Environmental Analysts Incorporated l' A- 224 Seventh Street 11530

• Garden City, tiew York l

g,,

i l

January 1973 \

l b

'4 l' ,-

t . ~ ~ - - - - --- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

. . , _ _ _ _ _ . _ 1 1

9 ABSTRACT _

Pilgrim Unit 1, a 1,998-Mut boiling water reactor station, began During this six-month start-up and initial cperation in June.1972.

testing period, reactor operation was intermittent and nower levels were quite variable. Radionuclide discharges in the liquid and gaseous ~

effluents were very low during this period. )

I The operational environmental radiation surveillance program is being conducted in accordance with the specifications.in the Pilgrim In addition to the environmental ronitoring i Station operating license.

f program outlined in the license, Boston Edison maintained surveillance  !'

No increases o- at other locations during this initial reporting period.

in environmental radioactivity levels outside the station property were hL The only statistically significant in-evident in the monitoring data.

crease in radiation levels during this period was th,e measured gamma at the overlook area TLD monitoring station, which is exposure rate /

t The measured

' located approximately 400 feet from the turbine building.

exposure rate increased approximately 10 microroentgens per hour above l

background at this location in November ' Access to this location is I visitors' center is controlled by Coston Edison Company; however, a small The average increcsed open to the public from April through November.

dose rate to the attendant was estimated to be from 0.6 to'0.8 millirem during this reporting period, and the average dose to individual visitcrs was estimated to be from 0.0004 to 0.003 millirem during October and November.

The population dose to visitors at the overlook area during .

4 4-

. x_.-- _ _ - - .._ -___ ._-.__ ____-_-_._L..__ - - _- - -

~ ~~ - -_

2 October and fiovember was estimated to be from 0.004 to 0.007 man-Based on these limited measurements the annual population dose to visitors fiore at the overlook area was estimated to be approximately 0.2 man-rem.

reliable dose estimates will be possible in the next semiannual report after additional measurements have been made.

C3 O

e o

e A. m.

3 i

I. INTRODtJCTION This is the first semiannual report of the Pilgrim Nuclear Power Station ope. rational environmental radioactivity monitoring program, covering the period from July 1,1972, to December 8,1972. The

- data are reported in accordance with the operational environmental monitoring and reporting requirements of Appendix A to the pilgrim Facility Operating License DPR-35, Technical Specifications and Bases.

Pilgrim Unit 1, a 1,998-MWt boiling water reactor, began initial start-up operations in June 1972.

Over this initial six-month period

.g -

the reactor power level was increased to approximately 98 percent of F

full pc'..cr in November. The gradual power ascension and frequent outages l

occurred as a result of performing various requira. tests of component  :

l systems. Figure i shcus the daily reactor thermal power levels frcm July through December. Radionuclide emissions in the liquid and gaseous effluents were minimal during this initi51 period of testing.

~

I l

The monitoring program has been exec,uted by Boston Edison Company personnel with analytical services provided by Interex Corporation (for- J merly ICN/Tracerlab).* Extensive measurements of background radiation f

were inade during the preoperational monitoring program and serve as a l baselire for evaluating the measured radioactivity levels during the .

c;crational program. The preoperational program began in 1969 and was also performed by Bosten U i :r personnel and Interex Corporation. Boston Edison perscrin:1 collected the environmental samples, except for marine _

life, and delivered them to Interex. Interex has performed the radioactivity l

L; '

l *1601 Trapelo Road. Waltham, Massachusetts.

I

. 4 l f analyses quarterly to treasurenents and has provided the resu ts o  :

Marine life samples have been collected by the Boston Edison.

Massachusetts Division of Marine Fisheries and provided to Bost Edison Cor.gany for analysis.

The thermoluminescent dosimeters (TLD) i a TLD ,

are changed monthly and read by Bos. ton Edison personnel us ng reader at Interex Corporation.

In addition to the monitoring program defined Tin the operating license technical specifications, Boston Edison operated four o locations air sampling stations and seventeen additional TLD monitoring This report includes the results obtained during this report period. i in the specified monitoring program and at the supplementary l l Im? rex Corporation had not completed all of the required ana The avail-

1. of environmental samples collected through December h sults 8, 1972.

able results are reported in Section III of this report, and t e re for incomplete analyses will be provided in the next semiannu i

D

. es t

10-N .

East Weynouth;(EU) 43 l

. Pilgrim Site

, ~

O- Dosimeter (TLD) ~,

Air Sampler and  ;

9 Dosirreter (TLD)

O miles I .

Cape Cod

[j Say

" A Duxbury fre3

' (SS O Kingston(KS)

O Bayshore llorth Drive (DD -

Rocky Hill Road (NR' Plymouthv.: , g Road Plymouth 1o_'i / ocky -

R Hill -(gp Center (PC)-

- South,PI) outh O ww ,7h # Manc.et(rd)

' '" ~* "

Cleft s la omet(ffsj OPly uth Airport Rock I' ** ~ -)

(SA) Area (CR

, ymouth (CP) Q J'

Sagamore (CS) 1: \ ,

't I- r n.., e 7,.

i,. 4, . .

r , ,; .

i:.g;ure and Air Particulate I;onitoring Stations.

1

- v.

O nosincter (ito) f Air San:pler and Dositeter (TLD)

Broken line = site boundary .

Sampling locations Preperty Linc I' .:estrian J and F. are not shczn Leidge Public Parkin.9 Area (PA) l'roperty Line ') L .[ .

1 Pilgrim Station- L CAPE C00 BAY l'ropert ,O ,' r y East Brea ater g Line (! I .,8 (EC) Otroperty k

ser Eo,k' '- jine(C) r; % ' , 'roperti.

Line(F ; Area (0A) r **

Line(H g w

• i Q

~

s roperk.' Ro k i YRockyllil \

Road (A) \

Eitts {! Road (B)

' ' ~ - p Prcperty \

-- L Line (11) \ -

lD rcperty Lin

)

IMS l i

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'k N k f')

O Microwave Tower (rtT)

/

'es p

Is s Property O(

• Line (Wil) , l N i l

l \ # 1 J'

' \ (~ I

\  !

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N

% I L

• I Property Q, Line (E) l Fie,i:re 3.

L: :atirns of Onsite Exposure and Air Particulate Monitoring Stations. {

. = - -

m .> _

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• 3 G "5 M @ M er = N 0 W N N 03 O ==

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ed MNWWm-N 68 e m er == N M N C% "6. S N Ps N. . 0 0. .,3O.000000

+ 3, . O. . . 0 0 0 e.

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C0000000000w0000000 6 f 80 E f .

so b C C=g eq. g O - M - c c - m m er N e e m m N

&C ==. 0. . -. - . O. . O. O. O. O. O. O. O. O. O. O. O. O.

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es C000000000000000000 b  !

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.at D ==

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e oo CD v -- O C - N c.ll c3 er e m m m e N n

• , Cs Cs . O. . . . =. =. 0. O. O. O. O. O. O. O. O. C. e='* O.

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==

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• 2- 3=3 7 3.3 3 3 as 3 C* ' . *' *'.; "".> aC. <..M. *C. <. M,e @O M. M.. O. C:l> O. O. O. I'.: C, gO o

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• = ----

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.x Table 7 3

Air Particulates--Gross Gansna Concentrations in Monthly Corposites (can/m )(a,b) 4 Control Offsite Onsite Collection Pedestrian East Rocky Property East Warel.ouse Period East Plymouth Cleft Manomet West Rocky Overlook Weymouth_ Center Rock Substation Hill Road _ Area Bridge 11111 Road I.ine Breatwater Sarpler (1972) 36 28 22 27 9 24 18 July 24 31 27 28 17 13 9 13 10 12 10 11 August 17 11 14 September 13 15 14 22 11 16 14 14 15 11 '(c) 7 6 5 6 5 4 5 5 OctoLer 5 1 5 Noveiber I d y g- - g- g g g g g g g

a. All values are to be multiplied by 10-3

b. Error is 226 or 10%. whichever is larger.

c. Instrument :nalfunction--no data available. j

d. I= incomplete analysis. Results will be reported in next semlannual report.

f u

M h

. - ~ . _- . .

i

.ev ,

i J

Table 8 .

Gaseous and Particulate Iodine-131 in Air Samples (pCI/m3 )(a)

Control Offsite Onsite Manomet West Rocky Overlook Pedestrian East Rocky Property East warehouse Collection Period East Plymouth Cleft Center Rock Substation 11111 Road Area Bridge liill Road Line Creatwater Sarvier (1972) lleyrouth

.)- <j ,y ,y ,y July 6-July 13 <1 <1 <1 c) <1 <1

= . = = .

= = =

July 13-esty 21 = = = = = = . .

=

July 21 -aly 27 = = = . . . . ..

=

July 2 7- . ..g . 3

= = = = = = = . . .

=-

Au J. 3- .g. 10 = = = = = = . .

= = = '

Auj. 10- 3 17 = = . . . . .

= =

Aoy. 17- .g. 24 = = = = = = . . ,

Aug. 24. , ..i . 30 =

= = " '

Aug. 30 ....t. 7 = " "

(b)

" = "

(b)

- = - --- = = =

(b)

5. r t. 1 'e : t. 14 " = =

<1

= = = = . . .

5. pt. 14-5. ;.t. 21 " " "

Sept. 21-Su t. 28 * = = = =

(b) (b)

" = = = = = <j = = ,y .

Sept. 28-Oct. 5 = = = = = =. . .

=

Oct. 5-Oct. 12

= = = = = = . = = .

=

Oct. 12-Oct. 19 " = = = = = = = = . .

Oct. 19-Oct. 26 = = = = = = = = = = .

Oct. 26-flov. 2 = = = = = = = = = .

Flov. 2-Ilow. 10 (b)' = = = = = = = =

flow. 10-flov. 16 <1

= =,

a. Both p.iLiculate lodine-131 concentrations on filters and gaseous fodine-131 concentrations on impregnated filters were less than minimum detect f.le concentrations in all samples. ('

b. Ins tri. nt malfunction--no data available.

ta N

I-l-

6

.sD^

,- PILGRIM NUCLE AR GENER ATING STATION l

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i Environmental Radiation Monitoring Program i

+

1 4

l:

l SEMIANNUAL REPORT NO.2 JANUARY 1,1973 THROUGH JUNE 30,1973 DATE OF ISSUE: AUGUST 29,1973 j

i-i i

r i

BOSTON EDISON COMPANY 4

9

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I. ABSTRACT S The operational environmental radiation surveillance program continues to be conducted in accordance with the specifications in Reactor operation during the the Pilgrim Station operating license.

Integrated radionu:lide i reporting period averaged 85% of full power.

l discharges in the gaseous effluent increased over the first semi-l

. annual period due to the increased plant capacity _ factor. Liquid J

effluents were maintained at low levels.by minimum liquid releases' and improved radwaste treatment.

A detailed post-operational radio

)

l

?

ation survey was made in January 1973 using a pressurized ion chamber l

and NaI spectrometer for comparison to a similar preoperational

(

survey conducted in July 1972.

No significant increases in environmental radiation and radio-activity levels beyond the station property were evident in the monitoring data, i f Onsite radiation levels, increased in areas close to the turbine g,,,,

building. The increases were attributed to direct radiation from ,

j l

the N-16 present in the turbine steam. The largest increase of any l

consequence (about 12 uR/h above background)-occurred at the over-look area, a visitors' area controlled by Boston Edison Company. 1 The average increased dose to the visitor population at this area I i-over natural background is estimated to be 0.08 man-rem during this reporting period. The dose accumulated by all of the visitor population to the Pilgrim Station (22,000 persons) is estimated at 3

0.1 man-rem during the reporting period. The guard at the public parking area, the two attendants employed there and at the overlook '

i area, and the four groundskeepers each received an estimated 0.6 ag.

2 mrem over the reporting period. The natural background exposure '

v

over this period was about 50 mrem.

f s The Cs-137 concentration in the seawater in the discharge canal was slightly higher than that of the intake canal although other reactor produced nuclides such as Mn-54, co-58, 60 and Zn-65 were The discharge levels of Cs-137 into the canal har not detectable.

been similarly as low as these other nuclides so that the slightly higher Cs-137 concentration may or may not be due to reactor operation. -

The bottom sediment in the discharge canal outfall has been.

sampled extensively and showed no significant. radioactivity, in particular no Co-58, 60 activity, as found in the discharge canal and reported in the previous report.

II. INTRODUCTION This is the second semiannuc1 report of the Pilgrim Nuclear Power Station operational environmental radioactivity monitoring program, 30, 1973. The covering the period from January 1, 1973 through June monitoring data are reported in accordance with the operational environmental monitoring and reporting requirements of Appendix A to the Pilgrim Facility Operating License DPR-35, Technical Specifica:-

tions and Bases.

Pilgrim Station began initial start-up operations in June 1972; however, approximately full power operation was not attained until During the current reporting period, gross thermal November 1972.

power generation averaged approximately 85 percent of full power operation. The radionuclide emissions in the gaseous effluents increased above those reported during the previous 6-month period due to increased power generation, however, the emissions remained .}

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Table 1 Air Particulates and Gaseous Radiciodine Surveillance Stations Distance and Sampling Location Direction from Station Offsite Stations EastWeymouth(EW) 23 miles NW Plymouth Center (PC) 4.5 miles W-WNW Manomet Substation (MS) 2.5 miles SE Cleft Rock Area (CR) 0.9 miles S b

Onsite Stations -

Rocky Hill Road (ER) 0.8 miles SE Rocky Hill Road (WR) 0.3 miles W-WNW;

Overlook Area (0A) 0.03 miles .

Onsite Stations Not Required by Operating License Property Line (PL) 0.34 miles NW Pedestrian Bridge (PB) 0.14 miles N EastBreakwater(EB) 0.35 miles ESE Warehouse (WS) 0.03 miles SSE 4 s

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

Y East Weymouth (Ul)'

.h) 1 l

15 mil

- Pilgrim Site .

O -Dosimeter (TLD) (,, ..

' Air Sampler and 3 Dosimeter (TLD) ,

yO miles

. i.

} ,

I Cape

. Cod y Bay Duxbury- *fles (SS O Kingston(KS) ,

9 ,

, , llorth O -

Plymouth (lip),

Plymouth _ ,

Center (PC) "d g Manc:' et(MP).

-~"~

. . . . South P1 .outh O f '1 _Y Man z.ct(ME) l

.outh Cleft ,O Ma cmet(f15)

- .i .

O P Airport Rock I: cmet(MB) l Area (CR,,  !

(SA)

)j l

ymouth (CP) O l

1 Sagamore (CS) l s

y .

9 Figure 1. Locations of Offsite Exposure and Air Particulate Monitoring Stations.

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Table 7 3 x 10' )(a)

Air Particulates--Gross Gama Concentrations in Monthly Composites (cpm Onsite Offsite East

• Rocky Hill Collection Control Rocky hill Property

• Overlook Pedestrian

• Plymouth Clef t Mantnet Bridge ___ Breakwater Rcad(East) Warehouse

• Period last Road (West) 1.ine _ Area _

Weymouth Center _ Rock Area _ Substation (197 L 9 10 13 7 5 8 5 9 7 7 7 6 January 6 17 12 6 5 6 2 7 February 2 9 8 <2 4 4 7 5 4 <2 5 9 8 6 March <2 6

<2 3 11 8 5 4 8 <2 42 April

<2 <2 <2

<2 42 <2

<2 <2 <2 May <2 5 <2 <2

<2 <2 3 3 2 4 3 3 T L 15 SP_

June -- - . 6 :n.n 2.!

a. Error is i 2 or 10 percent, whichever is larger. -

E -

i1 , 'I y,

• !!ot required by operating license. .

8 G

t

.. - ; mer .

O

= . .

Table 8

.. :t i1i Gaseous and Particulate Iodine-131 in Air Samples (pC1/m3 )(a)

Offsite i Onsite Collection Centrol Pedes trian* East

• Rocky Hill East Plymouth Cleft Manomet Rocky Hill Property

• Overlook Period Area _ Bridge _ Breakwater Road (East) Warehouse *

(1973) WeyTouth_ Center _ Rock Area Substation Road (West) tine

<1 <1 <1 ,1 ,j <j .i ,i ,j

=

Jan. 3-Jan. 11 . <1

=

<1

= = = = = =

=

=

Jan. 11-Jan. 18 = = = = = = = = = =

=

Jan.18-Jan. 25 " = * = = = = = = =

Jan. 25-Jan. 31 = = = = =. = = = =

=

Jan. 31-Feb. 8 = = * = = - = = = = =

Feb. 8-Feb. 15 = = = = =~ = = = = =

=

.Feb. 15-Feb. 22 = " * * = = = = = =

=

Feb. 22-Feb. 28 = = = " = = 4 =

Feb. 28-ftar. 8 " = = = = = = =

(b)

Par. 8-r.ar. 15 = = =

= = = = = ,j

=

=

ttar. 15-F.ar. 22 = = = = = = - = = =

=

Har. 22-Fiar. 28 " = = = = = .

(b)

=

l'.ar. 28-Apr. 5 '

=' ' =- - = = = ,j Apr. 5- Apr. 12 (b) = * = = - = = = = =

Apr. 12- Apr. 19 (b) " = = = = = .= = = =

Apr. 19-Apr. 26 <1 = = . =

" =1 *- = = . = - . - -

" -; a Apr. 26-Itay 3 * = 6 = = =

(b)

Kay 3-Itay 10 =

=

=

=

.:= ",

.j- ==

, , == =

- = - -. = = = ,g Hay 10-t:ay 16 = = = * .= = = = = , ,

r.ay 16-t:ay 24 " = = = = = = = = = =

F.ay 24-itay 11 = = = = = = = = = = =

! I?ay 31-Juna 7 = = = = =- = =

June 7-June 14

=

.= = = = = = = =

= =

Jur.e IC-June .21 = = = = = = =

= =

June 21-June 28 l

l l -

a. I-131 concentrations were less than the detection limit in all samples. .

J

b. flo sample. 1

• Not required by operating license.

M a

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4

^

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.-j. -

1972 Table A-7 Air Particulates--Gross Gama Concentrations in Monthly Composites (cpm /m )(*' }

f Offsite Onsite Control Pocky Hill Rocky Hill Overlook-East Plymouth Manomet Cleft Area Collection Period Substation Rock Road (West) Road (East)

(1972) Weymouth Center._

27(c,d) 36(c) 27(c) 28(c.d)

July 24(c) 31(c.d) 28(c.d) ,

13 10 9 11 17 14 ,

August 17 11 14 16 15 22 14 September 13 5 6(e) 5(e) 5(e)

October 5 1(e) 7(e) 4 3 5 2 <2 <2 <2 November 2 7 4 5 8 4 7 December

a. All values are to be multiplied by 10-3

b. Error is 2o or 10 percent.

c. Ru-103, Ru-106 Zr-95, and Nb-95 detected by gama spectrum analysis.

d. Ce-141 and Ce-144 detected by gama spectrum analysis.

e. Ru-103 and Ru-106 detected by gama spectrum analysis.

4

~~~ ~- - - _ _ , ~

fi y '

} .

{

p -

NJ- .

r i Tao.e A-8 ,

3({

Gaseous and Particulate Iodine-131 in Air Samples (pC1/m ) a Onsite Offsite _

Rocky Hill .0verlook; I

Control _ -Cleft Rocky Hill' Area _ _

Nanomet Road (East)

East Plymouth Rock ,. Road (West) _

Collection Period Weymouth Center _ _ Substation _- ~<1

<l'"

(1972)

<l 4

<1- "

<l <1 " " "

6-July 13 <1 " " "

July July 13-July 21 "

' July ~ 21-July 27 "

3 "

~ July 27-Aug. "

Aug. 3-Aug. 10 "

17 " "

Aug. 10-Aug. "

24  ; .

Aug. 17-Aug. "

(b) " ~" "

Aug. 24-Sept. 7 "

" <1 " "

Sept. 7-Sept. 14 "

Sept.14-Sept. 21 " "

" * "~ "

Sept. 21-Sept. 28 " " -

Sept. 28-Oct. 5 "

Oct. 5-Oct. 12 " "

Oct. 12-Oct. 19 " "

Oct. 19-Oct. 26 2

Oct. 26-Nov. "

Nov. 2-Hov. 10 (b) " "

16 <1 " "

Nov. 10-Nov. " "

Nov. 16-Nov. 22 "

" " "- i Nov. 22-Nov. 30 "

" (b) 7 " " " .1<

Nov. 30-Dec. "'

(b) "

Dec. 7-Dec. 15 "

" (b) "'

Dec. 15-Dec. 21 "

" " <1 Dec. 21-Dec. 228 "

Dec. 28-Jan. 4 l t ations on impregnated

. f a.

Both particulate iodine-131 concentrations on filters and gaseous iodine-131 c filters were less than minimum detectable concentrations in all samples.

b. Iristrument malfunction--no data available..  ;

i I

__ _____.____ _ _________ ______._________ _ _ _ ____________ _ . _ _ _ _ . _ _ . ~ . . _ _ _ _ _ _ - ..

TABLE D-2 SEMI-ANNUAL

SUMMARY

OF RADICACTIVE CASE 0US EFTLUENTS JANUARY - JWE.1973 APRIL HAY JUNE JAN. TEB. KAR.

Total Noble Cases (Curies) 2.02 E4 2.u4 E4 1.96 E6 1.65 E4 1.10 E4 1.52 E4 (a) Hain Stack 1,47 E2 7.30 El 5.39 E2 3.08 E2 (b) Reactor Building Vent 1.49 E2 1.37 E2 Total Kalongens (*-1) (Curies) 1.46 E-2 1.00 E-2 1.43 E-1 2.01 E-2 1.02 E-2 9.89 E-3 5.32 E-3 (a) Hain Stack 1.23 E 9.72 E-6 2.09 E-3 1.24 E-2 4.18 E-2 (b) Resetor Building vent Total Particulate Cross Beta-Gamma Radioactivity (*-1) (Curies) 8.98 E-4 1.07 E-3 8.12 E-4 1.31 E-3 (a) Hain Stack 1.20 E-4

• 8.26 E-3 1.72 E-4 1.63 E-4 1.19 E-4 1.01 E-4 2.68 E-4 6.57 E-4 (b) Reactor Building Vent .

Total Particulate Gross Alpha (*-2) (*-2) (*-2)

Radioactivity (Curies) (*-2) (*-2) (*-2) ,

Total Tritium (Curies) 2.19 E-1 1.37 E-1 2.14 E-1 1.94 E-1 1.25 E-1 1.52 E-1 1.26 EO (a) Hain Stack 1.39 EO 8.20 E-2 1.07 E0 1.81 E0 1.67 E0 (b) Reactor Building Vent 2-8 }-12 4-30 5-27 6-29 Haximum 24-Hr. Noble cas 1-17 1.09 E3 6.48 E2 8.45 E2 1.03 E3 1.56 E3 Release (Curies) 7.67 E2 Forcer.t of Applicable Limit for 2.33 3.15 4.44 3.14 2.50 1.87 Noble cases .

Fercent of Applicable Limit for 1.45 8.73 30.07 4.14 Ralogens & Particulates 1.15 1.06 Isotopes Released (Curies)

A. Ealogens 1.09 E-2, 1.67 E-2 2.24 E-2 1.85 E-1 2.54 E-2 Iodine-131 1.14 E-2 (*-3)

(*-3) 3.92 E-2 (*- 3) 5.41 E-2 Iodine-133 (*-3) (*-3)

(*-3) (*-3) (NDA*-4) (*-3) (NDA*-4)

Iodine-135

5. Particulates 2.21 E-5 3.51 E-5 4.66 E-4 3.10 E-4 3.87 E-4 4.79 E-4 Barium / Lanthanum-140 1.53 E-5 5.42 E-6 Beryllium-7 5.67 E-7 1.87 I-6 Cesium-134 8.20 E-6 7.49 E-6 8.56 E-6 1.63 E-5 8.16 E-6 Cesium-137 4.26 E-7 1.49 E-4 1.49 E-5 1.73 E-4 3.07 E-4 9.62 E-5 2.97 E-4 Chromium-51 3.82 E-5 3.78 E-5 4.72 E-6 2.65 E-5 1.lu E-4 Cobalt-58 3.49 E-6 2.99 E-6 1.99 E-6 1.06 E-5 3.22 E-6 Cobalt-60 3.56 E-6. 1.89 E-6 Iro n-59 2.32 E-6 1.96 E-5 6.21 E-6 Hanganes e-54 1.91 E-6 1.30 E-6 Zine-65 4.52 E-7 21rconium-95 1.52 E-7 Niobium-95 C. Cases (*-5) .

4.27 E2 3.84 E2 4.73 E2 .6.48 E2 8.12 E2 6.25 E2 Zenon-138 1.57 E3 1.88 E3 3.25 E3 4.17 E3 2.39 E3 Krypton-87 1.72 E3 2.96 E3 2.48 E3 1.77 E3 2.26 E3 3.57 E3 4.61 E3 Kryp ton-88 1.19 E3 1.56 E3 2.01 E3 1.37 E3 Krypton-853 1.05 E3 7.34 E2 2.80 E3 3.88 E3 6.12 E3 E.81 E3 5.43 E3 Xenon-135 4.09 E3 5.03 E3 2.72 E3 2.63'E3 4.16 E3 4.29 E3 6.34 E3 Xenon-133 1.36 E3 7.62 E2 1.64 E3 1.81 E3 4.02 E3 1.10 E3 Sus of Remainder

(*-1) With half-lives greater than 8 days

(*-2) Not measured since no alpha found in reactor coolant

(*-3) Quarterly analysis - Tech. Spec.

(*-4) NDA = No detectable activity

(*-5) Hain stack only O

h PILGRIM NUCLEAR GENERATING STATION D l Environmental Rartiation Monitoring Program

.~

.et g SEMIANNUAL REPORT FIO. 3 JULY 1,1973 THROUGH DECEMBER 31,1973 DATE OF ISSUE: M ARCH 1,1974 1

l .

BOSTON EDISON COMPANY l

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

9 I

i ABSTRACT radiation surveillance program continues The operational environmental to be conducted in accordance with the specifications in the Pilgrim Station operating license. Reactor operation during the reporting period

.' An administrative limit of 50% of full

' averaged about 60% of full power.

j power was maintained from October 6 until the maintenance and refueling shut-i Effluent discharges decreased as compared to the previous _

down on December 28. ,

semiannual period, primarily as a result of reduced plant capacity factor.

Detailed opertt*ing information, including a summary of plant releases, is reported l

. in a separate semiannual report entitled, " Operating and Maintenance Report."

f i

' No significant increases in environmental radiation or radioactivity

' levels beyond the station property were evident in the monitoring data.

l Radiation levels in areas close to the turbine building continued to show an increase over background. The increase was attributed to direct radiation g

l from the N-16 present in the turbine steem. Two onsite visitor areas were As a result, the increased radiation dose to the j subjected to this increase.

57,000 visitors over the reporting period was estimated to be 0.25 man-rem.

l .,

t I,ncreased exposures were also calculated for transient workers, working out-l side of the security area. Two visitor area attendants each received about l Four groundskeepers i

1.7 millirem (arem), while a third received about 0 9 area.

j absorbed approximately one millirem each and two research workers each received less than 0.4 mrem of increased radiation dose, over the reporting period.

+

a The background dose over this period was about 50 area.

6 g

m e- , - . , - - - - - _ _ - - - . _ _ . _ _ - -- _ . - -

k) East lleymouth (EW) _

Pilgrim Site O' Dostmeter (TLD) ~

4T Air Sampler and

.Desimeter(TLD) '

IO mileg

~

I Cape Cod

, Bay H

f;,3

) Kingston(KS)

Duxbury (55. O florth O Plymouth (lip)h.

Plymouth "

Center (PC) eM

- [7M

, # Man South P1 ,outh O l- /

l',c r. et(llE)

(SP) IO l Plymouth Cleft l' .st(115)

• l O Airport Rock A g. cmet(MB)

I j

(SA) Area (CRl ymouth (CP) O # l Sagamore (CS) f i

e Figure 1. Locations of Offsite Exposure and Air Particulate Monitoring Stations.

[ s

em i gi . Ja

%e o e. .V s

en sp en n sn se ess sn '

N @ re.* m eee= m r3 Ju r s e r N.= m * * .r e s es es 3 en e rJ.

o .

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4 C. O. e'1. e e. Ce G. **. e e. O.

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Air Partuculates--Gross Gamma Concentration in Fonthly Composites (cp/m ' ,

Onsite errsite East-Property

• osericcx Pcoc.tinr- P:c:y(hintC:stl "

Control P.ocky h m ReH cec *.e-:sai -

Collection Teit ;2nc- at Area Sriin. Brc & ater Ecst Plyacuth Line Pericd Veyr cuth Center _ Rock Arca Stb!tation Road (1:cr.t)_

(1973) 3 5 8 B N) 5 'l 6 5 4 July 4 10(b) 29 11 7 11 12 4 12 4 8 August 10 10 11 <2 4 4 <2 14 4 4

<2 <2 18 Septerber g 4 7 <2

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f i

PILGRIM NUCLEAR GENERATING STATION- ,,

Environmental Radiation Monitoring Program kt * .j

~ SEMIANNUAL REPORT NO. 4 JANUARY.1, 1974 THROUGH JUNE 30, 1974 l

l I ISSUED: i AUGUST 29, 1974 BY:

NUCLEAR ENGINEERING DEPT.

ENVIRONMENTAL SCIENCES GROUP

')

BOSTON EDISON COMPANY 9

4

_ --__1-_._ . _ - - - - . . . _ . _ - - - . - --

t I. INTRODUCTION AND

SUMMARY

i

.The Operational Environmental Radiation Surveillance Program '

continues.to.be conducted in accordance with the specifications of the Pilgrim Station Operating License. The reactor was shut down during the entire reporting period for refueling and mainte-nance and for contested licensing hearings regarding a change in fuel design.- As a result, gaseous effluent releases were very low and direct radiation from the facility was exceedingly small.

The gaseous Iodine -131 release for January, highest of the semi-annual period, was only about 1.5 percent of the release limit.*

• Due to-the increased in-plant maintenance activity, the liquid effluent releases were higher than during normal operation.**

This fact, coupled with intermittent operation of the circulating (dilution) water pumps, resulted in radionuclide buildup in the discharge canal. The highest levei of 5.2 pCi Co-60 per gram h sediment was detected in the clay found on the rocks at the sides of the canal. The Co-60 activity level decreased to 0.14 pCi/gm

- adjacent to the end of the canal and 0.05 pCi/gm 400 feet north-west of the canal. No activity was detected southeast of the f station (MDA = 0.01 pCi/gm).

-' The other media samples showed only radioactivity from natural sources and fallout from the Chinese nuclear weapon test of 4 December 1973. No plant-related nuclides were detected in these media.

t

• Radioactive effluent releases are summarized in the Appendix.

• • Details of plant operation can be found in a separate report

' entitled, " Operating and Maintenance Semiannual Report #4." ,

1

I II. DESCRIPTION OF MONITORING PROGRAM The Environmental Radiation and Radioactivity Surveillance Pro-gram is being performed in accordance with the requirements (DPR-35).

specified in the Pilgrim Facility Operating License Summaries.of the sampling media, locations, frequencies of  :

Details collection and analyses are given in Tables 1 through 4.

of the sampling program are given in Semiannual Report (2.*

Sampling locations are.shown in Figures 1, 2, and 3.

The radio-analysis of environmental samples is being performed by Interex Corporation. The limits of detection associated Samples of with

- their analytical procedures are given in Table 5.

bottom sediment and selected marine life were analyzed by The sensiti-Teledyne Isotopes using Ge(Li) gamma spectrometry.

vities of their analyses are given with the results in the appro-Nh priate tables.

.v  !

l.. ,

i 9

• Pilgrim Nuclear Generating Station, Environmental Radiation -

Monitoring Program, Semiannual Report #2, August 29, 1973.

9 2

a A ts J I

N l 'O h*lLQ

'O kileg ~

I Cape Cod M

(

1

• Mileg b

• P1ymouth Boy Y. 30 f

4 Y

- w 50 e p lg 9 *0

'10 0 14 0

11

' I O 12 p.

J I

P

)

r t

O DOSIMETER (TLD) '

f A AIR PARTICULATES AND 13 0 7

DOstMET ERS (TLD) p

" 3 - A LEGEND 9 MANOMET (ME) 1 DUX8URY (SS) 10 MANOMET IMS) 2 KINGSTON (KS) 11 MANOMET (MS) 3 NORTH PLYMOUTH (NP) 12 COLLEGE POND (CP) 4 PLYMOUTH CENTER (PC) 13 5AGAMORE (CS) 5 SOUTH PLYMOUTH (SP) 14 PLYMOUTH AIRPORT (SA) 6 BAYSHORE DRIVE (80) 15 EAST WEYMOUTH (EW) 7 CLEFT ROCK ARE A (CR) 8 MANOMET (MP) t

-l I

Figure 1. Location of Offsite Radiological Monitoring Stations 7 \

. . . . . - . ~ .

\

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a /

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SCALE IN FEET 4lA'+,bo, t a 22 et d ii a'.

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LEGEND l

13 ROCKY HILL ROAD (B) 1 PROPERTY LINE ID) 14 MICROWAVE TOWER (MT) 2 PROPERTY LINE (F)

" 15 EMERSON ROAD (EM) 3 PROPERTY LINE (1) 4 PROPERTY LINE (G) 16 WHITE HORSE RO AD (WH) 17 PROPERTY LINE (El 5 ROCKY HILL ROAD (A) 18 ROCKY HILL ROAD (WR) 6 PROPERTY LINE (H) 19 PROPERTY LINE UI 7 PUBLIC PARKING ARE A (PA) 8 PhDESTRIAN BRIDGE (PB) 2o PROPERTY LINE (K) 9 OVERLOOK ARE A (OA) 21 ROCKY HILL ROAD (ERI to E AST BRE AKWATER (EB) 22 PROPERTY LINE (L) 11 PROPERTY LINE (C) 23 WAREHOUSE (WS) 12 PROPERTY LINE (HB) 24 PROPERTY LINE (PL)

O DOSIMETER (TLD)

A AIR PARTICULATES AND OOSIMETERS (TLD) l l

1 l

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Figure 2. Location of Onsite Radiological Monitoring Stations j 8

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ty I i TABLE 7

~

)-

(*'

AIR PARTICULATES - GROSS GAMMA CONCENTRATION IN MONTHLY COMPOSITES (cpm /m3 x 10 l Onelta Offeito Bocky mill .

i

' . gM Cleft Menoset mocky Mill Property Overlook Podestrien Breaktsster Esot Road (Emet) Werehouse Collection Esot Phypouth Line Aree _ 3rldte Wynouth Center _ pock Area Subetetion Road (West)_

Period (1974) 14 11 12 I 10 9 15 25 12 4 11 7 Jon it is 14 It 14 4 I It 14 Feb 14 14 la 21 3'U8 16 le 16 13 12 15 11 12 12 Mar 21 12 26 23 22 24 I 19 1 Apr 11 I fa) Error le 12 or let, whichever le larger. 3 th) Cease spectrum indicates no nuclides identified greater than 0.05 pCi/m .

todientes incomplete snetystes results will be reported in meet semiennual report.

Ic) I M

44 b

h

_ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ __ . _ . - 4

7 TABLE 8

' PARTICULATE IODINE-131 IN AIR SAMPLES (pCi/m )

Control Offelte Onette Rocky Mall accky Mill East Flymouth Cleft Manonet mond most Property Overlook Fedestrian Fast Collection Period 1Anynowth Center Sock Area Subetation Onsite Line* Area Bridge' Breakwater * 'need OneiteEast unrehouse* .

1 Jan 3 - Jan 10 (a) <,1 (a) ta) (1 <1 tal <1 (s) <1 <1' Jan le - Jan 17 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 Jan 17 - Jan 24 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 Jan 24 - Jan 31 <1 <1 <1 41 <1 <1 <1 <1 <1 <1 <1 Jan 31 - Feb 7 <1 <1 <1 <1 <1 <1 <1 41 <1 cl <1 Feb 7 - F eb 14 <1 <1 <1 <1 41 <1 <1 <1 <1. <1 <1 Feb 14 - Feb 21 <1 <1 <1 <1 <1 <1 <1 <1 41 <1 <1 Feb 21 - Feb 29 <1 <1 <1 <1 <1 <1 <1 <1 (1 <1 <1 Feb 29 - Mar 7 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 Mar 7 - mar 14 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 Mar 14 - Imar 21 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 Mar 21 - Mar 20 <1 <1 <1 <1 <1 <1 <1 <1 41 <1 <1 Mar 28 - Apr 4 <1 <1 <1 41 <1 41 . <1 <1 <1 <1 <1 M Apr 4 - Apr 11 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 . tal A6 Apr 11 - Apr 10 <1- <1 <1 <1 <1 <1 41 <1 <1 <1 top Apr 18 - Apr 25 <1 c1 <1 <1 <1 41 <1 <1 41 <1 (a)

Apr 2$ - May 2 <1 <1 <1 <1 <1 41 <1 <1 41 <1 <1 may 2 - May 9 <1 <1 <1 <1 <1 <1 <1 <1 . <1 <1 <1 May 9 - par 16 <1 <1 <1 41 <1 <1 <1 <1. <1 <1 <1 May 16 - nay 23 <1 (1 <1 41 <1 <1 <1 <1 <1 <1 <1 May 23 - May 30 <1 <1 <1 <1 <1 <1 la) <1 <1 <1 <1 May 30 - June 6 <1 <1 <1 <1 41 <1 41 <1 <1 <1 <1 June 6 - June 13 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 (al No sample.

• 5tetton not requir54 by operating license.

l

yew ==, m -_ __

e i FE=r mmw =

TABLE 8A GASEOUS IODINE-131 IN AIR SAMPLES (pCi/m )

Onsite Control Offsite Docky nkil Rocky Mall Road East East Flymouth Cleft Manoanet Road Property Overlook Fedestrian East W eite Warehouse

• West Line Area . Drloge__ Breakwater

• Collection Period Weymouth center _ Rock Area Substation (a) <1 ta) <1 <1

<1 (a) (a) <1 <1 Jan 3 - Jan 10 (a) <1 <1 <1 <1 41 ta) (a) <1 <1 (a)

Jan IS - Jan 17 ta) <1 tal tal (a) (a) ta) ta)

Jan 17 - Jan 24 ta) (m) tal ta)

<1 <1 <1 41

<1 <1 <1 <1 <1 *1 *<1 <1 <1 <1 Jan 24 - Jan 31 <1 <1 <1 <1 <1 Jan 31 - Feb 7 <1 <1 <1

<1 <1 41 <1 <1

<1 <1 <1 <1 <1 <1 <1 <1 Feb 7 - Feb 14 <1 <1 <1 <1 *1 <1 Feb 14 = reb 21 <1 <1 <1 <1 <1 <1 <1

<1 <1 <1 <1 <1 <1 <1 Feb 21 - Feb 29 <1 <1 <1 <1 <1 <1 7 41 <1 <1 <1 <1 <1 <1 Feb 28 - Mar <1 <1 <1 *1 *1 Mar 7 - Mar 14 <1 <1 <1 *1 <1 <1 <1

<1 <1 <1 <1 <1 <1 <1 <1

<1 Mar 14 - Mar 21 <1 <1 <1 <1

<1 41 <1 <1 <1 <1

<1 <1 Mar 21 - Mar 28 <1 <1 <1 <1 <1 <1 Mar 29 - Apr 4 <1 <1 <1 <1 <1 tal

<1 <1 <1 <1 <3 <1 Apr 4 - Apr 11 <1 <1 <1 <1 <1 el ta)

<1 <1 <1 <1 <1 <1 <1 to)

M Apr 11 - Apr 18

<1 <1 <1 <1 <1 <1 <1 LA Apr 18 - Apr 25 <1 <1 (b) (b) (b) (b) (bl Apr 25 - May 2 (b) (b) (b) (b) (b) thi

<1 <1 <1

<1 <1 <1 <1 <1 <1 <1 <1 May 2 - May 9 <1 <1 <1 <1 <1 9 - May 16 <1 <1 <1 <1 <1 <1 <1 <1 May <1 1 <1 <1 <1 May 16 - May 23 <1 <1 <1 <1 <1 <1 <1

<1 <1 <1 <1 tal <1 May 23 - May 30 <1 <1 <1 <1 <1 <1

<1 <1 <1 <1 <1 <1 May 30 - June 7 <1 <1 <1 <1 <1 <1

<1 <1 51 <1 June 7 - June 13 <1 <1

<1 <1 <1 <1 <1 <1 <1 <1 <1 June 13 - June 20 <1 <1 (a) Mo sample, tbl Sanple toet during laboratory move.

• Station not required by operating 11eenee.

I<

TABLE A-1 .:

SEMIANNUAL

SUMMARY

OF RADIOACTIVE LIQUID EFFLUENTS JANUARY-JUNE, 1974 My June Total g g My Ay

1. Gross Radioactivity 1.08E0 7.54E-1 5.14E-1 2.76E-1 4 . 0 3E-1 3.79E0 (a) Total Released (C1) 6.7tE-1 3.4 3E-7 3.06E 7 2.71E-7 1.52E-7 2.07E-7 (b) Avg Cone Released ( C1/ml) 9.50E-8 3.12E-7 1.00E-6 2.20E-6 (pC1/al) 2.03E-7 2.20E-6 1.17E-6 1.10E-6 1.09E-6 (c) Man Cone Released

2. Tritium 1.90E0 1.01E0 5.46E-1 7.22E-1 1.03E1 3 (C1) 3.53E0 2.59E0 3.62?-7 (a) Total Released 7.55E-7 8.64E-1 7.59E-7 5.35E-7 2.28E-7 (b) Avg Cone Released ( C1/ml) 4.94E-7

3. Dissolved Noble Casee * * * * *

(a) Total Released (C1)

(b) Avg Conc Released ( C1/ml)

4. Crose 31pha Radioactivity' 3.00E-4 2.32E-4 5.92E-4 8.77E-3 1.02E-2 (a) Total Released (C1) 5.00E-5 2.75E-4 5.57E-10 (b) Avg Cone Released ( C1/ml) 3.17E-12 1.50E-11 1.64E-11 1.27E-11 5.00E-10 2.77E-9

5. Volume of Liquid waste to 1.5826 1.0126 6.5ES 1.11E6 0.50E6 Discharge Canal (Liters) 2.52E6 1.62E6 2.20E9 1.33E9 1.02E9 3.17E9 1.83E10

6. Volume of Dilution water (Litere) 7.14E9 3.43E9

7. Isotopes Released (C1) 6.51E-3 5.21E-2 1.27E-3 1.79E-3 9.79E-2 Chromium-51 3.02E-2 4.04E-2 2.97E-2 2.52E-1 3.53E-2 4.70E-2 4.60E-2 5.35E-2 ,

2.est-1 Manganese 54 1.17E-1 3.74E-2 2.60E-2 1.59E-2 Cobalt-58 4.2eE-2 4.86E-2 2.24E-2 s.97E-3 1.34E-2 1.15E-1 8.96E-1 tron 59 6.55E-2 1.15E-1 3.292-1 1.52E-1 1.19E-1 L cobalt-60 7.7tE-3 1.07E-2 7.6tE-3 1.22E-2 6.77E-3 4.92E-2 h Eir.c-65 Strontium-09 Strontium-90 4.03E-3 2.12E-3 2.77E-d 1.33E-3 3.08E-4 1.5sE-3 6.32E-4 1.52E-3 6.46E-4 2.70E-4 1.05E-4 5.40E-4 7.55E-5 6.44E-5 6.90E-3 2.03E-3 3.02E-3 1.36E-3 1.31E-4 1.07E-3 7.17E-4 Antimary-124 9.73E-2 7

lodine-131 9.07E-2 6.56E-3 6.47E-2 4.4sE-1 i

s.06E-2 1.78E-1 4.712-2 5.76E-2 2.04E-2 l Costum-134 6.16E-1 1.45E-1 1.82E-1 5.56E-2 2.20E-1 1.45E0 Cesium-137 2.27E-1 5.60E-5 Barium / Lanthanum-140 5.60s-5 4.17E-4 Neptunium-239 4.17E-4 2.20E-2 1.83E-1 o 9.01E-2 4.44E-2 4.03E-3 1.40E-2 1.35E-4

[ Unidentified

8. Percent of Tech Spec Limit for - 12.73 18.95 Total Activity Released ** - - 25.12 -

M

• no detectable activity

" Rased on 10 C1/ quarter limit El I

O e

A-1

TABLE A-2 1 8

SEMIANNUAL'

SUMMARY

OF RADIOACTIVE GASEOUS EFFLUENTS JANUARY-JUNE, 1974 i

Agr har June total Jan Feb Mar

1. Total poble Gases (C1) *-2 *-2 *-2 4.63E0

• -2 6.03E2 (a) Main Stack 4.63E0 *-2 1.22E2 1.22E2 8.44E1 6.37El .

(b) Reactor Su11 ding Vent 0.64E1 1.24E2

2. Total Melogene 8-1 (C1) *-2 *-2 *-2 3.65E-5 3.45E-5 *-2 *-2 *-3 2.83E-3 (a) Main Stack *-3 *-3 2.39E-3 4.292-4 1.04E-5 (b) Reactor Building Vent ~~

3. Total Particulate Crose beta- (C1)

Gamma Radioactivity 8-1 *-2 *-2 1.47E-6

• -2 *-2 1.47E-6 *-2 1.85E-4 6.99E-4 3.86E-3 (a) Main Stack 1.89E-3 4.49E-4 3.73E-4 2.66E-4 (b) Reactor Sullding vent

4. Total Particulate Crose Alpha 8-2 Radioactivity (C1) *-2 *-2 *-2 *-2 (e) Main stack 2.42E-8 *-4 6.12E-8 *-4 *-4 2.40E-7 *-4 (b) meactor Building Vent (C1) *-2 1.40E-2

5. Total tritium 1.40E-2 *-2 *-2 *-2 *-2 1.10E-1 2.25C-2 2.08E0 (a) Main Stack 4.96E-1 1.10E0 2.44E-1 1.00E-1 (b) Reactor Building vent 4-19 5-16 6-7 1-2 2-21 3-29 2.62

6. Kazimum 24-hr poble (Date) 5.14 4.56 4.00 3.37 Ges Releases (C1) 8.52

7. Percent of Applicable Limit f or 0.05 0.03 0.03 0.04 0.03 0.05 0.05 l Noble Cases

8. Percent of Applicable Limit for 0.17 0.11 0.44 0.70 2.62 0.60 0.23 0 Halogene and Particulates

9. Isotopes Released (C1) 2.87E-3 i A. Halogens *-3 *-3 *-3 2.43E-3 4.292 4 1.04E-5 *-4 I Iodine-131 *-3 *-4 *-4 *-3 *-4

• -4 todine-133 *-4 *-3 *-4

,,,/ *-3 *-4 )

Iodine-135 1

3. Particulates 3.27E 5 5.48E-5

• 2.21E-5 3.81E-4 8.78E-4  ;

Beryllium-7 8.09E-5 1.02E-4 6.08E-6 1.25E-4 Manganese-54 1.83E-4 2.90E-6 4.11E-5 2.99E-5 2.60E-4  ;

1.15E-4 3.47E-5 3.65E-5 0.66E-5 )

Cobalt-58 9.35E-6 2.04E-5 5.68E-5 1.00E-3 fron-59 1.44E-4 9.00E-6 2.12E-4 2.63E-4 2.79E-4 9.63E-5 7.69E-7 9.17E-6 4.75E-5 J Cobalt-60 1.06E-5 9.35E-6 1.18E-6 4.54E-5 Eine-65 1.64E-5 4.20E-6 1.11E-6 '

2.51E-5 0.67E-6 6.29E-6 1.42E-6 6.47E-6 sirconium-95 6.33E 5 )

Niobium-95 1.84E-6 2.89E-5 1.92E-5 7.90E-6 5.37E-7 l

~

Ruthenium 103 5.37E-1 Ruthenium-106 5.98E-7 6.11E-5 1.432-5 8.42E-5 1 Cesium-134 2.222-6 9.44E-5 6.10E-5 1.75E-4 1

3.21E-6 5.63E-6 3.57E-6 2.79E-6 9.93E-4

.' Cesium-137 herium-Lanthanum-140 9.93E-4 3.20E-5 2.50E-5 4.43E-6 2.55E-6 3.722 6 6.76E-5 3 Cerium-141 1.16E-5 1.19E-5 6.85E-6 2.73E-6 Cerium-144 3.08E-5 *-4 <6.27E 7 *-4 *-4

<1.76E-6 *-4 Strontium-89 42.69E-7 *-4 *-4 6.17E-7 *-4 *-4

• Strontium-90 *-2 *-2 *-2

• -2 *-2 *-2 C. Cases **5 e

• -1 With half-lives greater than 8 days. main etack out of service.

• -2 Unit out for refueling and maintenances I

• -3 No detectable activity. l

• -4 Quarterly analysis - Tech Spece. i

• -5 Main stack only. \

l 1

i 1

4 A-2

BOSTON EDISON COMPANY PILGRIM NUCLEAR GENERATING STATION Environmental Radiation Monitoring Program

~"

SEMIANNUAL REPORT NO. 5

' JULY 1, 1974 THROUGH DECEMBER 31, 1974 .,

4

-1 i

p Prepared By

- Joel I. Cehn

• Environmental Sciences Group I, Nuclear Engineering Department l

] ..

March 1974 1

'J 1 -

l l

Approved By: /

G. J s Davis, Manager l_

Envi nmental Sciences Group -

l

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

4. l L

I .I. INTRODUCTION AND

SUMMARY

~This" report describes the data accumulated in the Environmental i . Radiation Surveillance Program during the semiannual; period .

JulyL1 through December 31, 1974.

After a seven-month outage for refueling, maintenance, and ,

contested licensing proceedings, the unit came up to full power l

l i

on August 16. . Plant capacity factors (a measure of electrical output) during.the reporting period were: July, 5 percent',

f i August, 84 percent;. September, 73 percent; October, 83 percent;

. November, 82 percent; and December, 75 percent.*

f On-December 17, an augmented off-gas treatment system was put i

i into service. This system will effectively eliminate iodines, and greatly reduce the amount of radiogases released from the main stack. Also put into service was a liquid waste solidification

g. system, which is expected'to have the effect of significantly g

reducing the quantity of liquid wastes discharged to the bay. A

)

summary of radioactive effluents released during the reporting I l

period is presented in Appendix A.

l 5

l k'

t The liquid releases associated with and occuring during the

- seven-month outage (prior to completion of the solidification system) are evident in the program data reported in Section III of this report. Media in the vicinity of the station found with trace amounts of radioactivity from the station were Irish moss, f a

bottom sediment, mussels, lobster (1 individual), cod (2),

herring -(2) , bluefish (2) and cunner. A potential radiation dose g

to' humans consuming these media was calculated using " worst case" assumptions. This dose was calculated to be 0.26 millirem per year resulting from station produced radioactivity and 6.7

'* Details of plant operation can be found in a separate report -

entitled, " Operating and Maintenance Semiannual Report #5."

1

i l

1

\

These millirem from nuclear weapons test fallout radioactivity.

results may be compared to the natural background radiation levels No radioactivity of approximately 100 millirem per year.

attributable to station operation was detected in any terrestrial l media.

4 l;

Si i

m

• 7

.J

. 1 d ..

t

.a i 2

r

)

i A 1s .

I N

'* kilgg

'* *Ilts _-

~

capecod a

8*Y m 1

• 4/ Leg O

• • Mymme B*y  ;

' 30 ao

• w d

eO 0 14, to O

5 0 12 l

O DOSIMETER STLD) f

(

' i

& AIR PARTICULATES AND 13 O )

DOSf METERS (TLDi '

t s l

{

4

, j LEGEND 1 DUXBURY ($$) 9 MANOMET (ME:

' 10 MANOMET (MS) 2 KINGSTON (KS) 3 NORTH PLYMOUTH (NP) 11 MANOMET (M81 4 PLYMOUTH CENTER (PC) 12 COLLEGE POND (CPI 5 SOUTH PLYMOUTH (SPI 13 SAGAMORE (CSI 6 BAYSHORE DRIVE IBD) 14 PLYMOUTH AIRPORT (SA) 7 CLEFT ROCK ARE A (CR) 15 E AST WEYMOUTH (EW) 8 MANOMET (MPI l

Figure 1. Location of Offsite Radiological Monitoring Stations 9

f .

l I

l \

Y i p? '. ' f 1

8

\

,/, '

24 f[39A 8

1 f

T KALE IN FEET 0 000 1200 3

, ".. o A 11 o 1

23 re

.. . 4 !1,2_ _ .,. ,

L. 5 %___.__ , ,

13 * :- i; . y

%.r. g4*9 ..... p,[\

. g _21'4' g /

t-L ..i

\  :

i l 14 ') f

• .., .s, l15 I A -.

l [,a=~.,_~. I

, ,# 16 e E

O!

19

• • s 17 9.S N

'o I l F

- LEGEND 1 13 ROCKY HILL ROAD (B) 1 PROPERTY LINE (D) l 14 MICROW AVE TOWER (MT)

~

2 PROPERTY LINE (F)  ;

, 15 EMERSON ROAD (EMI l 3 PROPERTY LINE (1) 4 PROPERTY LINE (G) 16 WHITE HORSE ROAD (WHI II 17 PROPERTY LINE (El

5 ROCKY HILL RO AD (A)

" 18 ROCKY HILL ROAD (WR) 6 PROPERTY LINE (H) 7 PUBLIC PARKING AREA (PA) 19 PROPERTY LINE U)

~

B PEDESTRIAN BRIDGE (PB) 20 PROPERTY LINE (K) I 9 OVERLOOK AREA (OA) 21 ROCKY HILL ROAD (ERI  ;

'* 10 EAST BRE AKWATER (EB) 22 PROPERTY LINE (L) l 23 WAREHOUSE (WS) 11 PROPERTY LINE (C) 24 PROPERTY LINE (PL) 12 PROPERTY LINE (HB) f l

O DOSIMETER (TLD) l A AIR PARTICULATES AND DOSIMETERS (TLDI  !

-1 Figure 2. Location of Onsite Radiological Monitoring Stations 10 l

~

1 .

g .

i e

TABLE 7A AIR PARTICULATES - GAMMA ISOTOPE-CONCENTRATION IN MONTHLY COMPOSITES.

(pCi/m 3) (a) --

Control Offeite Onsite Collection East Plymouth Cleft Manomet Rocky Hill Property Overlook Pedestrian East Pocky Hill Period (1974) teeymouth center Rock Area substation Road (west) Line Area Bridge Dreakwater Road (East) teorehouse September R-40 0.10 +0.00 0.2 +0.2 0.09 +0.07 (b) (b) (b) (b) (b) tb) (b) 0.11 +0.07 Th-220 (b) 0.2 +0.2 0.16 +0.009 (b) (b) 0.012 +0.009 (b) (b) (b) (b) (b)

Ru-106 (b) (b) (b) (b) (b) (b) (b) (b) (b) (b) 0.0 3 + 0.0 3 October g R-40 (b) (b) (b) 0.00 +0.04 (b) (b) G.05 +0.05 (b) (b) (b) (b)

W 220 (b) (b) (b) (b) (b) (b) (b) (b) (b) (b) 0.009 +0.000 Re-106 (b) (b) (b) (b) (b) (b) (b) (b) 0.03 +0.01 0.04 +0.02 ,

(b)

(a) Results of Ge(Lil spectrometry. No isotopes other than those listed in table were detected. Analysis required quarterly.

(b) Lene than MDA (Minimum D tectable Activity). Typical 90A's are R-40 0.05 pel/m Th-220 0.01 Re-106 0.03 I

G

u .w . - - a p i __ - ,

TABLE 7B -

AIR PARTICULATES - STRONTIUM-90 CONCENTRATION IN QUARTERLY COMPOSITES (pCi/m3.x 10-4) (a) (b)

Control Offsite Onette Collection East Plymouth Cleft Manomet Rocky Hill Property Overlook Pedestrian East- Rocky Mill Period (1974) %seymouth Center Rock Area Substation Road (West) Line Area Bridge Breakwater Road (East) teorehomme Jan - Mar 2 +1 6 11 4 11 2 11 1 +1 4 11 el 2 11- 5 +4 e2 3 +1

,Apr - Jun 4 +1 7 +2 5 +1 2 +1 4 +2 6 11 6 +1 6 +1 3 +1 5 +1 3 +1 Jely - Sept (c) 7 +1 el 5 +1 e +1 3 11 6 +1 5 +1 3 11 3 il 3 1)

Oct - Oec 4 +1 2 +1 (c) 3 +1 5 +1 4 +1 1 +1 2 +1 3 +1 2 +1 - 2 41 (a) 3.0 in table means 3 x 10-4 rC1/m 3

-(b) These results are reported with an error corresponding to two standard deviations in the counting error. At the low count M rates encountered, it is dif ficult to verify the half-life of Y-90 and the actual error could be larger than that reported, m

(c) Sample lost in analysis.

i

+

-~

~

.. "', 1 O t ) , _1 '

) . i

~

TABLE 8 PARTICULATE IODINE-131 IN AIR SAMPLES (pCi/m ) L Control Offsite Onsite East Plymouth Cleft Manomet Rocky Hill Property Overlook Pedestrian- East, flocky Mill Collection Period weymouth Center pock Area Substation Read (West) Line* Area Bridge

• Breakwater

• Road (East) Marehoose*

June 13 - June 20 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 June 20 - June 27 <1 <1 <1 <1 <1 <1 <1 <1 (a) <1 <1 June 27 - July 4 <1 <1 <1 <1 <1 <1 (a) <1 (a)- <1 -<1 July 4 - July 11 <1 <1 <1 <1 <1 <1 <1 <1 (a) <1' <1 July 11 - July 11 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 July It - July 25 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 July 25 - Aug 1 <1 <1 <1 <1 <1 ,<1 <1 <1 <1 <1 <1 Aug 1 - Aug 8 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 Aug 9 - Aug 15 <1 <1 <1 <1 <1 <1 <1 .<1 <1 <1 <1-Aug 15 - Aug 22 <1 (a) <1 <1 <1 <1 <1 <1 <1 <1 <1 Aug 22 - Aug 29 <1 (a) <1 <1 <1 <1 <1 <1 <1 <1 <1 Aug 29 - Sept 5 <1 (a) <1 <1 <1 <1 <1 <1 .<1 <1 ' <1-Sept 5 - Sept 12 <1 <1 <1 .<1 <1 <1 <1 <1 <1 <1 <1 Sept 12 - Sept 19 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 Sept 19 - Sept 26 <1 (a) <1 <1 <1 <1 <1 <1 <1 <1 <1 +

Sept 26 - Oct 3 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 g <1 <1 (a) <1 <1 ' <1 q Oct 3 - Oct 10 <1 <1 <1 <1 <1 oct to - Oct 17 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 Oct 17 - Oct 24 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 Oct 24 - Oct 31 <1 <1 <1 <1 <1 <1 <1 <1 (a) <1 <1 Oct 31 - Nov 7 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 Nov 7 - Nov 14 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 Nov 14 - Nov 21 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 Nov ~21 - Nov 28 <1 <1 <1 <1 <1 (a) <1 <1 <1 <1 <1-Nov 20 - Dec 5 <1 <1 <1 <1 <1 (a) <1 <1 . <1 <1 <1 Dec 5 - Dec 12 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1.

l Dec 12 - Dec 19 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 Dec 19 - Dec 26 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 <1 Dec 26 - Jan 2 <1 <1 <1 <1 <1 <1 -<1 <1 <1 <1 <1 l

la) No sample.

• Station not required by operating license.

f 9

---

___._e _ - _ _ w - m,

~--a

• ** L--~ - =E '

lQ* --

TABLE 8A GASEOUS IODINE-131 IN AIR SAMPLES-(pCi/m )

Offeite Onette Control Rocky Mill Property Overlook Pedestrian East pocky Hill East Plymouth Cleft Manceet Center pock Area Substation Road (18est) Line* Area Bridge

• Breakwater

• Road (East) Warehouse

• Collection Period Weymouth

<1 <1 <1 <1 <1 <1 <1 June 20 - June 27 <1 <1 <1 <1 <1

<1 <1 <1 <1 <1 <1 <1 <1 June 27 - July 4 <1 <1 <1 <1

<1 <1 <1 <1 <1 <1 <1 <1 (a)

July 4 - July 11 <1 <1 <1 <1 (el <1 <1 July 11 - July 18 <1 <1 <1 <1 <1

<1 <1 <1 <1 <1 <1 (a) <1 July 18 - July 25 <1 <1 <1 <1

<1 <1 <3 <1 <1 ' <1 <1 <1 July 25 - Aug 1 <1 <1 <1 <1

<1 <1 <3 <1 <1 <1 <1 <1 Aug 1 - Au g S

<1 <1 <1 <1 <1 <1 <1 3 - Aug 15 <1 <1 <1 <1 Aug

<1 ~<1 <1 <1 <1 <1 <1 <1 <1 Aug 15 - Aug 22 <1 (a) <1 <1 <1 <1 <1 Aug 22 - Aug 29 <1 (a) <1 <1 <1 <1

<1 <1 <1 <1 <1 <1 <1 - <1 Aug 29 - sept 5 <1 tal <1 <1

<1 <1 <1 <1 <1 <1 <1 <1 Sept 5 - Sept 12 <1 <1 <1 <1

<1 <1 <1 <1 <1 <1 <1 <1 Sept 12 - Sept 19 <1 <1 <1 <1 <1

'<1 (a) <1 <1 <1 <1 <1 Sept 19 - Sept 26 <1 <1 <1 <1 <1 Sept 26 - Oct 3 <1 <1 <1 <1- <1 <1

<1 <1 <1 (a) <1 <1 <1 Oct 3 - Oct 10 <1 <1 <1 <1 <1-

<1 <1 <1 <1 <1 .<1 <1 Oct 10 - Oct 17 <1 <1 <1 <1 <1

<1 <1 <1 <1 <1 <1 <1 (1 DJ Oct 17 - Oct 24 <1 <1 <1

<1 <1 <1 <1 <1 <1 <1 <1 (a)

(D Oct 24 - Oct 31 <1 <1 <1 <1 <1 <1 <1 Oct 31 - Nov 7 <1 <1 <1 <1 <1

<1 <1 <1 <1 <1 <1 <1 Nov 7 - Nov 14 <1 <1 <1 <1

<1 <1 <1 <1 <1 <1 <1 <1 Nov 14 - Nov 21 <1 <1 <1 <1 <1

<1 <1 <1 <1 <1 (a) <1 <1 Nov 21 - Nov 20 <1 <1 <1 <1 Nov 20 - Dec 5 <1 <1 <1 <1 <1 (a) <1

<1 <1 si <1 <1 <1 <1 <1 <1 Dec 5 - Dec 12 <1 <1 <1 <1

<1 <1 <1 <1 <1 <1 <1 <1 <1 Dec 12 - Dec 19 <1 <1 <1 <1 <1 <1 <1 Dec 19 - Dec 26 <1 <1 <1 <1

<1 <1 <1 <1 <1 <1 <1 Dec 26 - Jan 2 <1 <1 <1 <1 (al no aample.

(b) Sample lost during laboratory move.

• Station not required by operating Itcense.

~r . .

w. ,e e .

m.

8 6 - - .4 e

1 a i 2

4 TABLE A-1 SEMI-ANNUAL

SUMMARY

OF RADIOACTIVE LIQUID EFFLUENTS JULY - DECEMBER 1974 July Aug. Sept. Oct. Nov. Dec. Total

1. Gross Radioactivity 4.34 E-1 (C1) 1.28 E-1 4.87 E-3 1.23 E-2 2.18 E-2 1.64 E-1 1.03 E-1 (a) Total Released 2.65 2-8 5.45 E-8 4.09 E-8 5.99 E-8 (b) Avg. Conc. Released (uci/ml) 4.32 E-7 1.78 E-S 3.7 3 E-8 1.91 E-7 9.78 E-8 8.24 E-7 (c) Max. Conc. Released (uC1/ml) 8.24 E-7 '3.17 E-8 1.56 E-7 1.35 E-7

2. Tritium 1.99 E-2 3.33 E-2 1.75 E-1 (a) Total Released (Ci) 1.15 E-1 2.42 E-4 6.71 E-5 6.75 E-3 *

(b) Avg. Conc. Released (uci/ml) 3.89 E-7 8.86 E-10 2.03 E-10 8.20 E-9 6.61 E-9 1.32 E-9 2.41 E-8

3. Dissolved Noble Gases (a)

(a) Total Released (Ci) (a) (a) (a) (a) (a) (a)

(b) Avg. Conc. Released (uCi/ml)

4. Gross Alpha Radioactivity (a) Total Released (C1) <7.05 E-5 <1.77 E-7 1.10 E-4. 3.8 3 E-7 <1.81 E-7 <2.13 E-7 <1.81 E-4 (b) Avg. Cone. Released (uci/ml) <2.38 E-10 <6.48 E-13 3.33 E-10 4.65 E-13 <6.01 E-14 -'8.45 E-14 <2.50 E-11

5. Volume of Liquid waste to Discharge Canal (Liters) 2.35 E*5 1.97 E+4 9.16 E+3 1.74 E*4 1.81 E+4 1.65 E44 3.16 E+5

6. Volume of Dilution water (Liters) 2.96 E+8 2.73 E+8 3.30 E+8 8.23 E+8 3.01 E+9 2.52 E+9 7.25 E+9 -

gy I 1. Isotopes Released (Ci) 2.32 E-2 b# Chromium-51 7.59 E-4 5.45 E-3 6.43 E-3 1.05 E-2 Manganese-54 1. 2 3 E- 2 1.04 E-3 1.71 E-3 1.10 E-3 4.11 E-3 2.58 E-3 2.30 E-2 4.11 E-3 1.18 E-4 1.64 E-3 3.36 E-4 1.14 E-3 5.90 E-4 7.93 E-3 Cobalt-58 6.66 E-4 3ron-59 3.67 E-5 3.05 E-4 2.40 E-4 8.42 E-5 4.54 E-2 1.98 E-3 3.84 E-3 2.04 E-3 2.77 E-3 3.40 E-3 5.94 E-2 Cobalt-60 6.52 E-3 zine-65 5.85 E-1 9.97 E-5 2.08 E-4 1.25 E-4 1.25 E-4 1.12 E-4 3.58 E-5 1.10 E-4 6.56 E-6 3.37 E-6 1. 0 3 E- 3 Eirconium/ Niobium-95 8.60 E-4 1.35 E-5 1.04 E-3 9.98 E-3 Molybdenum-99/Technecium-99m 8.94 E-3 1.01 E-4 2.18 E-5 5.42 E-5 1.83 E-5 1.95 E-4 Silver-130m l od i n.,- l i t 1.35 E-4 1.87 E-5 1.24 E-3 3.91 E-3 5.96 E-2 8.29 E-1 7.32 E-2 lodine-133 4.53 E-4 1.98 E-6 6.16 E-5 1.43 E-4 4.79 E-4 1.14 E-3 Cesium-134 1.38 E-2 2.90 E-4 3.62 E-4 1.59 E-3 1.82 E-2 1.68 E-2 5.10 E-2 Cesium-136 9.10 E-5 2.65 E-4 9.07 E-4 1.26 E-3 Cesium-137 4.18 E-2 1.05 E-3 9.65 E-4 4.21 E-3 4.41 E-2 4.06 E-2 1.33 E-1 Ba r ium/Lant hunum-140 9.38 E-4 3.63 E-4 3. 2 3 E- 3 8.79 E-4 5.41 E-3 Cerium-141 4.20 E-6 4.24 E-5 1.42 E-5 1.32 E-5 7.40 E-5 Cer ium- 14 4 7.84 E-4 1.98 E-5 8.04 E-4 Neptunium-239 9.01 E-4 1.96 E-4 4.79 E-5 7.58 E-4 4.27 E-4 2.33 E-3 Strontium-89 1.10 E-4 1.44 E-4 6.23 E-5 5.92 E-4 5.07 E-3 3.29 E-3 9.27 E-3 ,

Strontium-90 1.27 E-4 9.46 E-6 9.16 E-6 2.96 E-5 3.08 E-4 1.99 E-4 6.82 E-4 Unidentified 1.07 E-3 1.10 E-4 1.30 E-4 1.12 E-3 8.78 E-3 1.29 E-2 2.41 E-2

8. Percent of Tech. Spec. Limit For Total Activity Released (bl 1.45 2.89 2.17 (a) No Detectable Activity (b) Based on 10 Ci/ quarter limit

m . -

C f * . .

e TABLE A-2 SEMI-ANNUAL

SUMMARY

OF RADIOACTIVE GASEOUS EFFLUENTS JULY - DECEMBER 1974 Agu Sept. Now Dec Total JA O 3c

1. Total noble Games (CA) tal Main Stack R.31 E*) 2.31 E*4 7.53 E*4 1.20 E*5 2.50 E*5 6.77 E*4 5.37 E*5 (b) Reactor Building vent 4.81 E*! 5.11 te2 5.72 E*2 1.04 E+3 2.50 E*3 3.09 E*3 8.55 E*3

2. Total malogene *-1 (CAB tal noin Stack 6.41 E-4 4.27 E-2 9.58 E-2 2.14 E-1 0.38 E-1 9.65 E-2 1.29 E O tbl Reactor Building vent 1.51 E-6 6.26 E-4 7.60 E-3 7.01 E-3 5.38 E-2 E.33 E-2 1.53 E-1

3. Total Pertteulate Croce peta-Gasma Radioactivity *-1 (Cil .

tal main Stack 2.40 E-4 1.88 E-3 1.02 E-3 1.50 E-3 1.11 E-3 3.83 E-4 6.13 E-3 tbl Reactor Building vent 2.59 E-4 7.76 E-4 9.50 E-4 5.00 E-4 2.70 E-3 3.55 E-3 0.74 E-3

4. Total Particulate Gross Alpha medsometivity (Cl) tal Main Stack *-2 *-2 5.30 E-9 *-2 6.99 E-S *-2 tbl mesetor Duilding vent *-2 *-2 3.10 E-9 *-2 1.54 E7 *-2

5. Total Tritium Ect) tal noin Stack 1.70 E-2 1.75 E-1 1.02 E-1 2.10 E-1 3.29 E-1 3.73 E-1 1.25 E O tb) mesctor Duilding vent 2.50 E-2 3.e3 E-1 2.04 E-1 6.85 E-1 3.08 E-1 3.03 E O 4.63 E O

6. nacimum 24-hr noble (Datel 7-11 8-27 9-29 10-31 11-5 12-1 Cao nelease (Cil 4.03 E*2 1.46 E*3 6.20 E*) 7.53 to) 1.24 E*4 8.26 E*3 7 Percent of Applicable Limit 38 For noble Cases 0.21 3.64 11.02 18.20 39.56 11.57' 14.02 0

S. Percent of Applicable Limit bd for nalogens 6 Particulates 0.19 2.20 0.48 11.57 61.e4 56.09 23.25

9. Isotopes peleased (Cil

a. halogene t od a ne- 3 31 6.43 E-4 4.33 E-2 1.03 E-1 2.22 E-1 S.92 E-1 1.80 E-1 1.44 E O- -

todine-133 *-2 *-2 3.15 E-I *-2 *-2 0.09 E-1 lodane-135 *-2 *-2 S.06 E-1 *-2 *-2 1.06 E O B. Particulates Chromnum-51 1.35 E-5 3. 35 E Ma ng a nese- 54 4.52 E-5 2.69 E-5 3.15 E-5 9.85 E-6 S.01 E-5 1.02 E-5 1.13 E-4 Cobalt-54 S.54 E-6 3.34 E-6 2.99 E-5 5.75 E-6 4.76 E-5 Iron-59 1.13 E-5 1.13 E-5 Cobalt-60 1.01 E-4 5.23 E-5 7.79 E-5 2.74 E-5 7.75 E-5 4.04 E-5 3.77 E-4 Eine-65 4.20 E-6 4.74 E-7 4.67 E-6 Er-Nb-95 3.33 E-6 3.33 E-6 Silver-llom 2.60 E-6 2.68 E-6 Antamony-124 1.95 E-5 1.95 E-5 Ceesum-lle 4.54 E-6 2.39 E-6 2.49 E-5 3.58 E-5 6.76 E-5 Ceesum-137 2.14 E-5 1.95 E-5 2.75 E-5 1.26 E-5 7.56 E-5 1.12 E-4 2.68 E-4 Ce r a um- 139 1.94 E-3 1.94 E-3 Bar n um- Lant hunum-140 5.66 E-4 3.35 E-4 5.72 E-4 5.43 E-4 1.69 E-3 3.71 E-3 Cerium-let 2.96 E-1 7.43 E-6 1.52 E-5 2.29 E-5 Cerium-144 3.94 E-6 5.37 E-6 9.31 E-6 Strontium-89 *-2 *-2 2.52 E-4 *-2 6.32 E-4 *-2 Strontium-90 *-2 *-2 1.37 E-6 *-2 3.26 E-6 *-2 C. Cases *-3 Renon-130 2.90 E*l 3.18 Ee 3 3.10 E+4 1,68 E*4 2.74 E*4 5.27 E*) 6.30 E*4 Erypton-87 2.29 E*2 2.97 E.) 3.45 E*) 1.61 E*4 3.49 E*4 0,52 Ee 3 7.12 E*4 Brypton-SS 2.37 E*2 2.50 E*3 1.14 E*4 2.07 E*4 4.50 E*4 1.10 E*4 9.08 E*4 mrypton-Osm 0.08 Eel 9.04 E*2 3.91 E*3 7.45 E*3 1.91 E*4 4.7) E*3 3.63 E*4 zennn- 135 3.20 E*2 3.87 E+3 1.71 E*4 3.19 E*4 7.82 E*4 2.17 F*4 1.53 E*5 Renon-133 6.59 E*1 2.70 E*3 1.28 E*4 1.04 E*4 3.58 E*4 1.16 E*4 0.14 E*4 Sum of memanader 3.40 E*2 6.90 E*) 3.06 E*4 9.45 E*) 9.60 E*3 4.80 E*3 4.18 E*4 31 with h7aUlsves greater than 9 +9 eye

• -2 overterly an.ilyske-Terh. spece.

• =1 M. tin 9thh naty. k

/

BOSTON EDISON COMPANY PILGRIM NUCLEAR GENERATING STATION Environmental Radiation Monitoring Program SEMIANNUAL REPORT NO. 6 JANUARY 1, 1975 THROUGH JUNE 30, 1975

.)

1 J

Prepared By .

Joel I. Cehn a

Environmental Sciences Group Nuclear Engineering Department 7 September 1975 3

o Approved By:

G. J s Davis, Manager Envi nmental Sciences Group .

~

6 I. INTRODUCTION AND

SUMMARY

This report describes the data accumulated in the Environmental Radiation Surveillance Program during the semiannual period January 1-through June 30, 1975.

Plant' operation'during the reporting period is detailed in a separate report entitled "Operat ing an d Maintenance Semiannual I

~ Report No. 6". ' Power levels during this period were limited, r

!,1 administratively, to 80 percent of full power from February through May and 70 percent of full power during June. These

t-limits on power level were set by Boston Edison Company for reasons of operation and maintenance. Plant capacity factors (a measure of electrical output) during the reporting period were:

gg January, 36 percent; February, 34 percent; March 76 percent; th April, 42 percent; May, 55 percent; and June, 65 percent.

7..

t L The Environmental Radiation Surveillance Program was implemented

, I'

- during the reporting period as required by the Nuclear Regulatory L

Commission. Plant-related radioactivity levels decreased from 1974 levels in molluscs, algae, and sediment. The current levels 6.

of Mn-54, Co-60 and Cs-137 in these media are less than 1 picoeurie

. per gram.

Weapons test fallout was detected in air particulates, specifically Nb-95, Ce-144 and Cs-137. These elements were also detected in

.various other media. Gaseous radio-iodine was detected onsite at 1

levels less than 1 picoeurie per cubic meter. This I-131 was attributed to releases from the reactor building vent.

~

Finally, marine media data from 1974 were reviewed and calcula- ,

tions showed.that the seafood ingestion dose to an individual was less than 1 millirem per year.

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TABLE 7 -

AIR PARTICULATES - GROSS GAMMA CONCENTRATION IN MONTHLY COMPOSITES (cpm /m x 10 ) I"'

~

t Offsite Onsite Control Collection East Plymouth Cleft Manomet Rocky Hill Property Overlook Pedestrian East Rocky Hill Period (1975) Weymouth Center Rock Area Substation Road (West) Line Area Bridge Breakwater Road (East) Warehouse 3 4 10 6 4 7 Jan 5 6 <3 3 9 7 14 8 16 12 32 Feb 19 19 17 14 23 28 20 25 19 19 8 12 Mar 21 15 17 14 26 22 30 25 28 36 22 Apr 15 26 15 22 11 11 15 16 9 8 May 12 15 16 14 12 g

14 15 18 16 6 14 21 9 '20 June 20 13 (a) Error is +2 or 10%, whichever is larger.

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TABLE 7A AIR PARTICULATES - GAMMA ISOTOPE CONCENTRATION IN MONTHLY COMPOSITES (pci/m )(*)3 o.eite ca.troi offeit.

Pedeetrian Best Rocky eint Rocky Nitt Property Overtook Road toest)_ Wereheese Cettecties Eeet Ftymouth Cneft Menomet Doed (weet) Line ._ Aree eridge __ Sreakeeter Weymouth center _. Rock area Substation Period (1975)

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TABLE A-2 SEMI-ANNUAL

SUMMARY

OF RADIOACTIVE GASEOUS EFFLUENTS.

JANUARY - JUNE 1975 g June Total Jan Feb Mar MJa I

1. Total Noble Cases (Ci) 9.87 E*3 5.37 E+3 9.13 E*2 2.30 E+3 3.38 E*3 4.03 E+3 2.59 E*4 r (a) Main Stack 1.15 E+3 3.27 E+3 3.97 E+3 4.37 E+3 5.30 E+3 2.01 E+4

)

(b) Reactor Building vent 2.06 E+3

2. Total Halogens (a) 3.35 E-1 3.12 E-2 1.88 E-2 7.95 E-2 4.31 E-2 3.27 E-2 5.40 E-1 (a) Main Stack 1.09 E-1 5.54 E-2 6.62 E-2 1.26 E-1 1.63 E-1 1.13 E-1 6. 3 3 E-1 (b) Reactor Building vent

3. Total Particulates (a) (Cil 4.01 E-4 4.61 E-4 5.94 E-4 1.04 E-3 1.82 E-3 4.24 E-3 8.56 E-3 (a) Main Stack 1.10 E-3 1.72 E-3 2.62 E-3 3.37 E-3 3.81'E-3 5.21 E-3 1.78 E-2 (b) Reactor Building Vent

4. Total Particulate Gross

-Alpha Radioactivity (Cl) 1.51 E-7 (b)

(a) Main Stack (b) (b) 2.16 E-9 (b)

(b) (b) <6.59 E-8 (b) <1.75 E-8 (b)

(b) Reactor Building Vent i 5. Total Tritium (Cil 1.21 E+0 1.34 E+0: 5.02 E+0 (a) Main Stack 5.60 E-1 5.70 E-1 8.30 E-1 5.10 E-1 8.53 E+0 3.15 E+1 (b) Reactor Building Vent 3.40 E+0 3.81 E+0 5.91 E+0 5.75 E+0 4.11 E+0 1-26 2-17 3-26 4-10 5-3 6-30

6. Maximum 24-hr Noble Gas (Date) 4.57 E+2 5.26 E+2 Release .

2.35 E+3 1.15 E*3 5.03 E+2 1.08 E+3

7. Percent of Applicable Limit 1.95 l

{ for Noble Cases 2.24 1.37 1.23 1.89 2.12 2.64

! w S. Percent of Applicable Limit l

i for Halogens and Particulates 77.46 39.76 42.69 84.23 (c). 12.82 i

9. Isotopes Released (Cil .

A. Halogens 2.06 E-1 1.46 E-1 1.17 E*0 lodine-131 4.44 E-1 8.66 E-2 8.50 E-2 2.06 E-1 ~

(b) 7.75 E-2 (b) .1.96 E-1 (b) (b)

Iodine-133 (b) (b)

Iodine-135 (b) 1.35 E-1 (b) 2.39 E-1

3. Particulates - - 2.88 E-5 ,

Chromium-51 2.88 E-5 - - -

3.66 E-5 2.19 E-5 1.39 E-5 1.40 E-5 7.51 E-6 5.4 8 E 9.93 E-5 Manganese-54 6.15 E-6 7.74 E-5 ,

Cobalt-58 6.18 E-5 7.56 E-6 5.06 E-7 1.37 E -

- - - - 1.17 E-5 '

1.17 E-5 -

Iron-59 6.45 E-5 4.80 E-5 2.89 E-5 3.51 E-5 2.95 E-5 8.57 E-6 1.74 E-4 Cobalt-60 6.22 E-6 9.21 E-6 2.24 E-6 1.61 E-7 - 5.84 E-7 -

Eire /Niob-95 - - - 1.13 E-4 1.13 E-4 Silver-110m - -

3.59 E-5 1.94 E-4 Cesium-134 5.79 E-5 2.96 E-5 5.17 E-6 2.47 E-5 2.87 E-5.

2.63 E-6 - - - - 2.63 E-6 Cesium-136 -

8.99 E-5 1.19 E-4 1.03 E-4 6.73 E-4 Cesium-137 1.54 E-4 1.61 E-4 4.71 E-5 1.87 E-3 2.99 E-3 4.08 E-3 5.23 E-3 8.96 E-3 2.42 E-2 Barium-Lanth.-140 1.06 E-3 2.09 E-4 7.74 E-4 Cerium-141 2.27 E-5 3.59 E-5 1.31 E-4 1.56 E-4 2.19 E-4 2.40 E-5 - 2.40 E-5 Cerium-144 Strontium-89 (b) (b) 1.58 E-3 (b) 2.04 E-3 - -

Strontium-90 (b) (b) 7.37 E-6 (b) 1.17 E-5 - -

C. Cases (d)

(a) with lalf-lives greater than 8 days.

(b) Quarterly analysis - Tech. Specs. 't (c) Tech. Spec. change to quarterly limit.

(d) Unable to predict due to Augmented off-Gas System.

t

. M -

. - .-- r

• _ v . . - < ~ - -

- _ . - . __ _ _ - . _ _ . _ - + -

I i

ED5 TON EDIEDN COMPANY #

/~

Gs=sna6 Dec. css eco Sov6svo= 5'asse  ;

Bosvow. MassAcMustvTs 02199 March 1, 1976 l

$.' , , }. {

Director R:gfon I, Inspection and Enforcement U. S. Nuclear Regulatory Commission -

631 Park Avenue -

l King of Prussia, Pennsylvania 19406 Docket No. 50-293 f/ License No. DPR-35 Semi-Annual Report No. 7 Q Environmental Radiation Monitor:ng Program Gzntlemen:

In accordance with Pilgrim Station, Technical Speci.fication 6.9.C.2, we are hsreby submitting our seventh Semi-Annual Report. 41 separate report covering i

-hnical Specification 6.9.C.1 and entitled, " Semiannual Summary of Radio-

.tive Effluents" has been sent under separate cover.

Very truly yours, 1

Original Signed by G. Carl Andognini

~

Manager -

I Nuclear Operations

! cc: Director '

Office of Inspection and Enforcement f

) U. S. Nuclear Regulatory Commission l Washington, D. C. 20555 (20) i Director Office of Management Information j and Frogram Control l U. S. Nuclear Regulatory Commission Washington, D. C. 20555 (2) l

)i l

I l

-- - - - _ - - - - - - _ _ - __ . A

4 BOSTON EDISON COMPANY PILGRIM NUCLEAR GENERATING STATION Environmental Radiation Monitoring Program h SEMIANNUAL REPORT NO. 7 '"

JULY 1, 1975 THROUGH DECEMBER 31, 1975

=0 Prepared By Joel I. Cehn Environmental Sciences Group Nuclear Engineering Department

~

March 1976 k

r i

f h

I Approved By:

G. Ja Envie%'sDavis, Manager Group nmental Sciences

^

. . - . _ _ . _ _ . _ . . _ . - _ _ . _ _ _ . . _ . _ _ . _ _ . . _ _ . _ _ _ . _ . . _ = _ _ .

?

I. ' INTRODUCTION'AND

SUMMARY

i This report discusses the data accumulated by the Environmental Radiation Surveillance Program during the semiannual period July 1 through December 31, 1975. '

Pilgrim Station operation during the reporting period is detailed:

in a separate report entitled " Operating and Maintenance Annual l

. 4 Report, 1975." Power levels'during thf.s period were limitad, t

administratively, to 70 percent of full power from July through ij

" October, and to 60 percent of full power during November and o

December. These limits on power levels were set by Boston Edison 1 .

9 Plant capacity Company for reasons of operation and maintenance.

h l' factors (a measure of electrical output as a percentage of full j

>- output 100 percent of the time) during the reporting period were:

j,.

~

July, 43 percent; August, 53 percent; September, 19 percent; q' October, 3 percent; November, 50 percent; and December, 53 per-b h

cent.

j)

The Environmental Radiation Surveillance Program was implemented l' I b during the reporting period as required by the Nuclear Regulatory

. Commission. Plant-related radioactivity levels in marine life

- and sediments decreased from 1974 levels, as did liquid effluent i

releases to Cape Cod Bay. The current levels of Mn-54, Co-60, Zn-65 and Cs-137 in these media are less than 1 picocurie per l

gram.

Gaseous radioiodine was detected onsite at levels less than 1 picoeurie per cubic meter. This I-131 was attributed to releases .

' from the reactor building vent.

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TABLE 7

! AIR PARTICULATES - GROSS GAMMA CONCENTRATION IN MONTH'.Y COMPOSITES (cpm /m x 10-3) (a) l Control offaite 'Onsite Collection East Plymouth Cleft Manomet Rocky Hill Property Overlook Pedestrian East Itocky Hill Period (1975) Weymouth Center Rock Area Substation Road (West) Line Area Bridge Breakwater Road (East) Warehouse 9 8 9 13 6 3 3 9 9 5 4 July August 4 7 <3 9 8 <3 6 7 7 11 7 September 5 <3 5 7 9 4 <3 <3 14 8 <3 October 11 9 7 8 4 5 20 7 4 <3 E M Movember 4 <3 <3 5 <3 6 8 <3 <3 <3 8 b

December <3 <3 12 <3 <3 <3 <3 <3 3 <3 10 (a) Analytical error is +2 or lot, whichever is larger.

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i TABLE 7B

' AIR PARTICULATES - STRONTIUM-90 CONCENTRATION IN QUARTERLY COMPOSITES-(pCi/m x 10-4)

Offsite Onsite Control Rocky Hill Property Overlook Pedestrian East pocky Hill Collection East Plymouth Cleft Manomet Period (1975) Weymouth Center Rock Area Substation Road (West) Line Area Bridge Breakwater Road (East) Warehouse 14 1 4 13 + 2

'Jan - Mar 13 1 2 <2 <2 12 1 2 812 11 1 4 12 1 2 11 1 2 13 1 3

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8', ,

EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT -(1975) r GASEOUS EFFLUENTS-ELEVATED RELEASE CONTINUOUS MODE D ATCH MODE Quarter Quarter Ovarter ,

Urds Ouarter - ,

Nuctides Released

1. Fission gues

.-

• E E Ci . E . E . .

k rypton.85 ~E E Ci . E . E . .

k rypton.M5m E E E krypton.57 Ci . li _ _,. . .

E E . E

• Ci E . .

krypton SS .

E E Ci . li . E . .

xenon.133 E E . E Ci E . .

li xendn.135 li Li .

Ci E . .

'sinon.135m Ci . li . E . li E

.senon.138 E E . E Ci E . .

Ottiers (>pecify) .

E E -

Ci . E . E .

n_ . _ _E .. ...._ e_- _E E t-

.....m... ..

E E

Ci 1. 74L4 1.3 0 E+ & .

'. unidentitied -

Ci 1. 7 4Ev. 1 30Ek. , . E . .E lotal for period

~ -

2. lodines 4 .4 0 E-1 1.4 0 E-2 E E MI 'I f M "I iodinc.131 ~Cl' E E Ci 5 19 E-l 2 5 3 E-2 . .

iodine.133 Ci 5 7 8. E-1 a. 44E2 . E . E (, lOl iodine.135 E E

Ci 1 53 E!O 2 07 E-1

- Total for period ,

-w- .

- (g~ a'

3. Partreufates .

E E sirontiurn.S9 Ci 9 . 3 2 E-3 6 . I 9 E-3 . .

E E 5 . 3 5 E-5 3 54 h-5 strontium 40 Ci .

E E 1 7 5 E. 5 6 80 E-f. * -

cesiunt.134 Ci Ci 2 .13 E 4 1 2 9 E.4 . E li cesiuni 137 E E ba riusu lantignuin.l.10 Ci 2 . 0 0 E--2 1. a. 0 E-21 . .

E E rannr7 nege'.7. ;1t < Cl 1.8 6 E-5 8 4 6 E-G . .

Ci- 6 .12 li-7 .nr 3 . li . E

~

enN1+_sp

• li li

.t. 7 !!<. <> pt .

tine-(s' Ci .

3 E

• E

.Ci 3 .g o E-5 3 81 li 5 -

c nb l +._'F 5 zirconium-niobium-95 ci 3;087,-5 c,857 -7 cerium-141 C1 2,64 E-5 7,21 J ~6 cerfun-lhh Ci 3. 2 7 F.C < !'DA nepuniun-239 Ci 1,3y 5 <: .y unidentified C3 Total for veriod C$ 2,55h-2 2 , 0 'i l. -2 i E.g s Pace 3 or 9

a- ..,,a -

..g. ,a..

, EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT" (1975)

GASEOUS EFFLUENT.S-GROUND-LEVEL RELEASES

m. 8- CONTINUOUS MODE BATCH MODG

. Nucikles Released Urdt Quarter Quarter Quarter Quarter .

I. Fission pses kr>gon 85 Cl . E . E . E . E

. k rypton-85 m Ci . li . li . E li krypton-87 Ci . E . 11 . li . E I.rvpton.Sx Cs . li . li li . I:

xeno.s.lI3 Ci . li . li . li . li xenED3 Ci . E . E . E F xenon.135m . Ci . 11 . E . 11 . E

.. xenon.138 - C E . E , E li

_ . Others (specify) Cs . E . E . E . li-

--.- Ci . E . E . li li q m. . . .- O . li . 11 . E . li unidentified Ci 1. S S E t. 1 3 311 L E . E Total for period ,

Ci 1. S S E 8 1 33E 4 . E . EI s

.

• 2. lodines -

- iodine.131 Ci 4.92E-l 7. 3 5 E -2 E F C' C' I ~ \

iot l inc.133 Ci 6.551i-11 2.421!-) . E . E T. D 6 ~ L ind:ne :.'S O C. C9h-1 :. 3 &b-1l . ti i . L. i 1. 3 ') F C 1 Total for period Ca 2 08 E40 1 725+61 1 . h I . E

b. i

3. Particulates i

I st rontium-89 Ci 6. 5 7 E-3 5 6 2 E-3 . E . E strontiumM0 Ci 3 .12 li-5 2 .9 9 li- S li E '

cesium.l.tl Ci ) .3 6 li-4 7 16 E -5 . E . E-cesium 137 C 1. 9 2 LA 5 0 5 E-4 . li . li '

barium lantlanum.i.10 Ci 4 .92 li-1 7 .3 5 E-2 . E li

,1...rr 43, El Ci 7 . 5 5 li- 5 <!'DA E . E C 1. 61 E-4 1. 02 li-4 . 11 . li vn t rnp h , "'g e e ).

Ci 2 11 li G 7.521i-6 . li . li l jynn 59 0 <MDA }.73 Li-6

. E

• E j zine-C5 Ci < bm.a 3.3GE-5 col.nlt-60 C' < 2.13h 1 i. rSE i. I zirconium-niotium-c5 ci 5.00r-5 s.55E-5 zilver-110n ci 2. I 8E-3 1.44E-3 ce riut .-) h ) ci 8. 13 E- te 1.20E-8 cerfur.-1kh Ci <MDA 4.03E-6 4

Total rot- veried ci 5. 0 2 E -1 a . 2 0 E -7 i .

i Page h of 9 l

l I

1 l

EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1975)

-- GASEOUS EFFLUENTS-SUMMATION OF ALL RELEASES Unit Quarter Ouarter Est. Tota 3 1 E rror. %

A. Fission & actitution pses ci 3 .2 9 Ee 2.63I!# G .50 Ffj

1. Total release pCi/sec 4 .14 13T 2 .31 liF3,

2. Averne release rate lor seriott

3. l'encent of Teeloneal specilic:nion linut 9 2 .52 E+0 2 .32 Etol

. . . - II. lodines Ci 9. 32r-1 l 8. 7 519 6. 0 0lii!

1. Total lodine 131

2. Avera;te scle.:5cratrfagviii.sl pCi'see 1.1717 1.101:-2

~

, l'~s7I:tl" 7. 2 0lF ,

3. l'encent of t 'ifiliU5lI[veltidtiiore linen

~

C. Partictilates Ci 5 . 3 2 E-3 1. 0 2 t?-1 6.50111 !

1. Particulates with lialf. lives >H ilavs ) . 2 6 li-2,

2. Averaec release-rate lor petiod pCi/see G . 6 9 h-2

3. Percent of technical s;weiliention linut 9 1. 4 9El 2 . 4 6 EKl

4. Gross alpha radioactmlv Ci 5 76 E4 8 . 2 7 li 7l D. Tritium Ci 2.48 E 1 1.31 E 1 5 . 0 0 E+1

1. Total release 3.12!! 0 1.65 E O

2. Average release rate for period uCi/sce

'4 . li . E

3. l'eereat af iachnirmin't. . cili. atinn iimit, I-t t

l 4

• e

$4 1%f;c 5_of 9

EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1975)

GASEOUS EFFLUENTS-SUMMATION OF ALL RELEASES Unit Quarter Quarter Est Tonar 3 4 E rror. %

A. rission & actlation gases Ci 3 .2 9 E4 2 .G 3 IIe 6 . 50 Ffj

l. Total release

2. Average release rate tor genod pCs/see 4.34l!d. 2 .3) liF3

3. l'e: cent of leelnucal speciti:ation liniit  % 2 .52 190 2 .3 2 11+0

' * . II lodines i

Ci 9. 32 r-1 1 8. 7 519 6. 0 0l!P.

1. Total indine.131

2. Avesye ici..:re rate f.w (v,rigd uCi/E)),._1717[3.3Oli

'. . 5118 l,l. 201.'0 ,

3. Pricent of teeint: cal spectuention futut .

C. Particulates t

Ci 5 . 3 2 E-1 1.02II-l 6. 5 0 til .!

1. Particulates with half lives >H davs

2. Averace release rate for pened pCi/sce 6 .6 9g2 1. 2 6 li-2,

3. Percent of technical s;veuicanon lintit 'A 1. 4 9 E+1 2 . 4 6 lie)

4. Gross alplia radioactivity Ci 5 78 E 4 8 2711-7)

D.Trillum

. Ci 2 . te 8 E 1 1. 31 E 1 5 . 0 0 l!+1

1. Total release

2. Average release rate for period _ , ,

pci/see 3.12!! 0 1.65!! O

'4 I;

• b y, pneeriWrT GiW@eiti asion limit -

i e

?

t m  %

t=

\ ,

Tu(;c 5 of 9

BOSTON EDISON COMPANY i

PILGRIM NUCLEAR GENERATING' STATION ,

Environmental Radiation Monitoring Program SEMIANNUAL REPORT NO. 8 JANUARY 1, 1976 THROUGH JUNE 30, 1976 e

N0

/' ' Prepared By I*

Joel I. Cehn ,

r ,.,,

Environmental Sciences Group l

Nuclear Engineering Department i

\

f August 1976 L.

'I Approved'By: -

G. Ja s Davis, Manager Envi nmental Sciences Group i

I. INTRODUCTION AND

SUMMARY

This report discusses the data accumulated by the Environmental Radiation Monitoring Program during the period January 1, through June 30, 1976.

Pilgrim Station operated at 19% of capacity during January and 53% of capacity during June. The unit was shut down for refueling and maintenance from January 29 to June 4.

The Environmen'.al Radiation Monitoring Program was implemented during the reporting period as required by the Nuclear Regulatory r ea Commission. Plant-related radioactivity levels in marine life

"" and sediments increased from 1975 levels, as did liquid effluent

(~. The predominant plant-related radio-

,. releases to Cape Cod Bay.

nuclides detected in these media were Cs-137, Co-60 and Mn-54.

'l Activity levels of these nuclides did not exceed 1 picocurie per gram of sample. The dose equivalent associated with the maximum activity concentrations detected was calculated to be less than 1 millirem per year for the seafood ingestion pathway.

'~

The only plant-related activity detected in terrestrial media was gaseous radiciodine. This was detected onsite at levels less than 1 picoeurie per cubic meter. This radiciodine was attributed to short-term releases which followed the uncovering of the reactor for refueling.

1

TABLE 6 AIR PARTICULATES - GROSS BETA CONCENTRATIONS IN WEEKLY SAMPLES (pCi/m )

East Plymouth Cleft Manomet Rocky Hill Property Overlook Pedestrian East Rocky Hill Period (1976) Weymouth Center Rock Area Substation Road (West) Line Area Bridge Breakwater Road ( East ) Warehouse Dec 30 - Jan - 7 0.02 0.02 0.02 0.03 0.03 0.02 0.02' O.03 0.04 0.04 0.05 Jan 7 - Jan 14 0.03 0.03 0.04 0.01 0.03 0.03 0.02 0.02 0.02 0.03 0.03 Jan 14 - Jan 22 0.05 0.05 0.05 0.04 0.05 (a) 0.04 0.04 (a) 0.05 0.04 Jan 22 - Jan 29 0.04 0.04 0.03 0.05 0.04 .0.03(b) 0.03 0.05 0.03(b) 0.04 0.03 Jan 29 - Feb 5 0.03 0.05 0.04 0.05 0.05 0.03 0.03 0.04 0.04 0.04 0.03 Feb 5 - Feb 12 0.04 0.04 0.03 0.03 0.03 0.03 0.03 0.03 0.05 0.04 0.03 Feb 12 - Feb 19 0.04 0.05 0.04 0.05 0.05 0.05 0.04 0.05 0.05 0.05 0.05(c)

Feb 19 - Feb 26 0.05 0.04 0.04 0.03 0.04 0.03 0.04 0.04 0.05 0.04 0.04 Feb 26 - Mar 4 0.03 0.06 0.02 0.02 0.04 0.03 0.04 0.03 0.02 0.03 0.04

~

Mar 4 - Mar 11 0.02 0.02 0.05(c) 0.03 0.02 0.03 0.03 0.03 0.05 0.02 0.04 Mar 11 - Mar 18 0.01 0.02 (d) 0.03 0.02 0.02 0.02 0.02 0.02 0.03 0.02 Mar 18 - Mar 25 0.02 0.03 0.03 0.03 0.03 0.03 0.03 0.05~ 0.03 0.03 0.04 Mar 25 - Apr 1 0.02 0.02 0.02 0.02 0.02 0.02 0.02 0.02. 0.02 0.03 0.04 Apr 1 - Apr 8 0.03 0.03 0.02 0.03 0.03 0.03 0.03- 0.03 0.04 0.03 0.03 PJ 8 - Apr 15 Apr 0.03 0.03 0.04 0.05 0.03 (d) 0.03 0.03 0.05 0.03 0.04 Apr 15 - Apr 22 0.04 0.05 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 Apr 22 - Apr 29 0.02 0.02 0.02 0.01 ,0.02 0.02 0.02 0.02 0.02 0.01 0.02 Apr 29 - May 6 0.03 0.03 0.03 0.02 0.03 0.02 0.02 0.02 0.03 0.02 0.03 May 6 - May 13 0.03 0.03 0.03 0.03 0.03 0.03 0.03 0.03 0.03 0.03 0.03 May 13 - May 20 0.02 0.02 0.02 0.01 0.02 0.02 0.02 0.02(c) 0.01 0.01 0.01 May 20 - May 27 <0.01 0.01 <0.01 <0.01 0.01 <0.01 <0.01 0.01 0.01 0.01 0.02 May 27 - June 3 0.02 0.02 0.02 0.02 0.03 0.02 0.02 0.02 0.01 0.02 0.02 June 3 - June 10 0.03 0.03 0.02 0.02 0.02 0.02 0.02 0.03 0.03 0.02 0.03 June 10 - June 16 0.03 0.03 0.02 0.03 0.04 (a) <0.01(c) 0.03 0.03 0.05(c) 0.03 June 16 - June 24 0.02 0.02 0.01 0.01 0.02 0.02(b) 0.01 0.01 0.01 0.02 0.02 June 24 - July 1 0.03 0.02 0.02 0.03 0.03 0.03 0. t - ' O.03 0.02 0.03 0.03 (a) Instrument station inaccessible (b) Two week sample (c) Not full week sample *

(d) Instrument malfunction j ,

i

i TABLE 7:

i .-- AIR PARTICULATES - GROSS GAMMA CONCENTRATION -IN MONTHLY COMPOSITES J (cpm /m3'x-.'10-3)<(a) l Control Offsite Onsite Collection East Plymouth Cleft Manomet Rocky Hill Property Overlook Pedestrian East . Rocky Hill Period (1976) Weymouth Center Rock Area Substation Road (West) Line Area Bridge' Breakwater .. Road (East) Warehouse.

January <3 <3 <3 <3 <3 <3 <3 <3 <3 .c3 <3 February 10 6 11 5 11 11 ~4 4 'll C 14 March <3 6 <3 3 <3 <3 10 ' 9- 6- 15 4 April <3 <3 <3 <3 <3 <3 <3 <3 '<3. <3 <3 May 4 <3 10 4 6 3 <3 <3- 4 <3 7 M

h June 4 8 6 5 7 <3 <3 7 5 7 7' (a) Typical analytical error is + 3x 10'3 cpm /m 3 In the counter used, 1 cpm corresponds to 2.2 pCi of Cs-137.

I o

A.

T U

~

L._.

~ '

TABLE 7A AIR PARTICULATES - GAMMA ISOTOPE CONCENTRATION IN MONTHLY COMPOSITES (pCi/m )

3 I^) .

t Control Offsite Onsite East Plymouth Cleft Manomet Rocky Hill Property Overlook Pedestrian East Rocky Hill Collection Bridge ' Breakwater Road (East) Warehouse Period (1976) Weymouth Center Rock Area Substation Road (West) Line Area i Janua ry

<.007 <.007 <.000 <.008 <.907 <.000 c.006 c.006 Cs-137 <.008 c.008 <.008 ,

Mn-54 <.008 <.008 <.007 <.007 <.008 c.009 <.009 c.007 <.009 c.007 <.007 Co-58 <.01 c.01 c.01 c.009 <.01 <.01 <.01 <.009 c.01 <.009 c.009 Co-60 c.008 <.009 <.009 <.007 <.008 <.008 c.008 c.007 c.009 c.007 c.007 q

2n-65 c.02 <.02 <.02 c.02 <.02 <.02 c.02 <.02 c.02 c.01 c.02 1

bJ K-40 <.09 <.1 col <.1 .08+.05

<.07 c.2 <.07 .22+.09 col <.06 tn Be-7 .081 06 <.1 .081 07 .15+.06 .141 06 <.1 <.1 <.1 .161 08 .06+.05

.121 06 April Cs-137 4.006 < .000 4.008 <.007 <.008 <.01 < . 008 <.007 4.008 <.008 c.008 Mn-54 <.006 <.009 < .009 <.006 (.008 <.01 <.008 <.007 <.008 c.000 <.008 ,

r I

Co-58 <.008 <.01 <.01 <.009 < .01 <.01 <.01 4.009 <.009 <.009 <.01 Co-60 4.007 <.008 c.009 <.007 <.008 (.01 <.008 c.007 c.008 <.007 c.008 zn-65 <.02 <.02 <.02 <.02 <.02 <.02 <.02 <.01 < .0 2 <.02 <.02 K-40 <.1 <.1 < .1 <.09 <.1 <.02 .21+.09 .071 06 <.1 <.1 c ol ,

Be-7 <.08 <.1 .30+.009 .08+.08 <.1 .2+.1 <.1 .10+.08 .1+.1 .20+.09 .10+.10 Others .013+.009 Th-228 *

(a) Results of Ge(Li) spectrometry. Analysis required quarterly

g..

TABLE 8 PARTICULATE IODINE-131 IN AIR SAMPLES (pCi/m )

JANUARY - JUNE 1976 Offsite Onsite Control East Plymouth Cleft Ma nomet Rocky Hill Property Overlook Pedestrian East Rocky Hill Collection Period Weymouth Center Rock Area Substation Road (West) Line Area' Bridge Breakwater Road (East) Warehouse

<0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 Dec 30 - Jan 7 <0.02 <0.02 <0.03

<0.05 <0.05 <0.04 <0.04 7 - Jan 14 <0.04 <0.04 <0.06 <0.04 <0.04 <0.05 <0.05 Jan (a) <0.03 <0.03 Jan 14 - Jan 22 <0.03 <0.03 <0.03 <0.03 <0.03 (a) <0.03 <0.03

<0.02 <0.01(b) <0.02 <0.02 <0.01(b) <0.02 <0.02 Jan 22 - Jan 29 <0.02 <0.02 <0.02 <0.02

<0.02 <0.02 <0.02 <0.02 <0.02 Jan 29 - Feb 5 <0.01 <0.02 <0.02 <0.02 <0.02 <0.02

<0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 Feb 5 - Feb 12 <0.04 <0.04 <0.03 < 0.03 (c)

<0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 Feb 12 - Feb 19 <0.02 <0.02 (0.02 <0.02 <0.02 <0.02 Feb 19 - Feb 26 <0.02 <0.02 <0.02 <0.02 <0.02 4 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 Feb 26 - Mar

<0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 Mar 4 - Mar 11 <0.04 <0.04 <0.15(c) <0.04

<0.04 <0.04 <0.04 bJ Mar 11 - Mar 18 <0.04 <0.04 (d) <0.04 <0.04 <0.04 <0.05(c) <0.04 0% <0.02 40.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 Mar 18 - Mar 25 <0.02 <0.02 <0.02

<0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 Mar 25 - Apr 1 <0.02 1 - Apr 9 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 Apr

<0.02 <0.02 <0.02 (d) <0.02 <0.02 <0.02 <0.02 <0.02 Apr 9 - Apr 15 <0.02 -0.02 <0.02

<0.02 <0.02 <0.02 <0.02 0.0310.02 <0.02 <0.02 <0.02 <0.03 <0.02 Apr 15 - Apr 22 <0.02 <0.02 <0.02 <0.02 <0.02 Apr 22 - Apr 29 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02

<0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 Apr 29 - May 6 <0.02 <0.02 <0.02 <0.02

<0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 May 6 - May 13 <0.02 <0.02 <0.02

<0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.04(c) <0.02 May 13 - May 20 <0.04 <0.04 <0.04 May 20 - May 27 <0.04 0.04 <0.04 <0.04 <0.04 <0.04 <0.05 <0.04

<0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 <0.02 May 27 - June 3 <0.02 <0.02 <0.02

<0.02 <0.02 <0.02 <0.03 <0.02 <0.02 <0.02 <0.02 June 3 - June 10 <0.02 <0.02 <0.02

<0.03 June 10 - June 16 0.03 <0.03 <0.03 <0.03 <0.03 (a) <0.04 <0.03 <0.0) <0.09(c)

<0.03 <0.03 <0.03 <0.02(b) <0.03 <0.03 <0.03 <0.03 <0.03 June 16 - June 24 <0.03 <0.03

<0.03 <0.03 <0.03 <0.03 <0.03 <0.03 <0.03 <0.03 <0.03 <0.03 June 24 - July 1 <0.03 (a) Instrument Station inaccesible (b) Two week sample (c) Not full week nample (d) Instrument malfunction

u m .L l l 3 O T l ;

TABLE 8A GASEOUS IODINE-131 IN AIR SAMPLES (pCi/m )

JANUARY - JUNE 1976.

Control Offsite Onsite

~

East Plymouth Cleft Manomet Rocky Hill Property Overlook Pedestrian East Rocky Hill Collection Period Weymouth Center Rock Area Substation Road (West) Line Area Bridge Breakwater Road (East) Warehouse Dec 30 - Jan 7 <0.02 <0.03 <0.04 <0.03 <0.02 <0.02 <0.03 <0.02 0.04!0.02 <0.02 0.04 0.02 Jan 7 - Jan 14 <0.04 <0.04 <0.06 <0.04 <0.04 <0.05 <0.04 <0.05 <0.05 <0.04 <0.04 Jan 14 - Jan 22 <0.04 <0.04 <0.04 <0.04 <0.04 (a) <0.04 <0.04 (a) <0.04 <0.04 Jan 22 - Jan 29 <0.05 <0.05 <0.05 <0.05 <0.05 <0.02(b) <0.05 0.0610.03 <0.02(b) <0.05 <0.05 Jan 29 - Feb 5 <0.03 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 0.0410.02 0.0410.02 Feb 5 - Feb 12 <0.04 <0.04 <0.04 < 0. 0 4 ' <0.04 <0.04 <0.04 0.0610.03 0.0610.03 - <0.04 0.1610.03 Feb 12 - Feb 19 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 0.0510.03 0.0510.03 <0.04 0.0710.04(c)

Feb 19 - Feb 26 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 Feb 26 - Mar 4 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 bJ Mar 4 - Mar 11 <0.04 <0.05 <0.17(c) <0.04 <0.05 <0.04 <0.05 <0.04 <0.05 <0.04 <0.04

-J Mar 11 - Mar 18 <0.05 <0.04 (d) <0.04 <0.04 <0.04 <0.06(c) <0.05 <0.05 <0.04 <0.04 Mar 18 - Mar 25 <0.05 <0.04 <0.05 <0.04 <0.04 <0.04 0.0510.03 <0.04 <0.05 <0.04 <0.04 Mar 25 - Apr 1 <0.04 0.0610.0 3 < 0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 0.0410.03 Apr 1 - Apr 8 <0.05 <0.05 <0.05 <0.05 <0.05 <0.04 <0.05 <0.05 <0.06(c) <0.05 <0.04 Apr 8 - Apr 15 <0.04 <0.05 <0.04 <0.04 <0.05 (d) <0.05 <0.04 <0.05 <0.04 <0.04 Apr 15 - Apr 22 <0.04 <0.05 <0.05 <0.05 <0.05 <0.04 <0.05 <0.05 <0.05 <0.05 <0.05 Apr 22 - Apr 29 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 Apr 29 - May 6 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.04 May 6 - May 13 <0.05 <0.05 <0.05 <0.05 <0.05 <0.05 <0.05 <0.05 <0.05 <0.05 <0.05 May 13 - May 20 0.0510.0 3 < 0.04 <0.04 <0.04 <0.04 <0.04 <0.04 <0.07(c) <0.04 <0.04 <0.04 May 20 - May 27 <0.04 <0.04 <0.04 <0.04 <0.05 <0.04 <0.05 <0.04 <0.04 <0.04 <0.04 May 27 - June 3 <0.05 <0.05 <0.05 <0.05 <0.05 <0.04 <0.05 <0.05 <0.05 <0.04 <0.05 June 3 - June 10 <0.05 <0.05 <0.05 <0.05 * <0.05 <0.04 <0.05 <0.05 <0.05 <0.05 <0.05 June 10 - June 16 <0.06 <0.06 <0.06 <0.06 <0.06 (a) <0.07(c) <0.06 <0.06 <0.18(c) <0.06 June 16 - June 24 <0.06 <0.06 <0.06 <0.06 <0.06 <0.03(b) <0.06 <0.06 <0.06 <0.06 <0.06 June 24 - July 1 <0.07 <0.07 <0.07 <0.07 <0.07 <0.06 <0.07 <0.07 <0.07 <0.07 <0.06 (a) Instrument station inaccessible 1 (b) Two week sample j (c) Not full week sample J (d) Instrument malfunction l

i l

TABLE-A-2 SEMI-ANNUAL

SUMMARY

OF. RADIOACTIVE GASEOUS EFFLUENTS 1976 Quarter 1 Quarter 2 Gross Radioactivity 1.14E 4 2.02E 4 Total Released (Ci) 1.45E 3 2.57E 3 Avg. Release Rate (uCi/sec)

Tritium 9.33E O 3.48E O Total Released (Ci) 1.19E O 4.43E-1 Avg. Release Rate (uCi/sec)

Gross Alpha Radioactivity (Ci) 2.62E-6 1.33E-6 Isotopes Released (all Curies)

Halogens-Main Stack 2.10E-2 3.61E-3 Iodine-131 3.29E-2 1.50E-2

'I Iodine-133 4.46E-2 2.16E-2

~ Iodine-135

e. Halogens-Reactor Bldg. Vent Iodine-131 1.89E-1 6.09E-3 Iodine-133 5.86E-2 2.11E-2 Iodine-135 1.05E-1 3.69E-2 l-Particulates-Main Stack Strontium-89 4.74E-3 6.57E-3 Strontium-90 3.50E-5 3.84E-5 Cesium-134 1.10E-5 5.46E-6 Cesium-137 1.31E-4 9.89E-5 Barium-Lanthanum-140 4.79E-3 2.10E-2 Manganese-54 2.02E-5 9.70E-6 Cobalt-58 2.63E-6 2.44E-6 Cobalt-60 4.10E-5 3.66E-5 Zirconium-Niobium-95 5.10E-6 4.74E-6 Cerium-141 4.04E-6 2.29E-6 Cerium-144 2.13E-6 1.75E-6 Iron-59 4.22E-6 1.07E-8 Zinc-65 2.33E-6 NDA I

Particulates-Reactor Bldg. Vent Strontium-89 1.38E-3 2.42E-3 .

Strontium-90 7.75E-6 4.65E-6 I Cesium-134 2.31E-4 1.07E-4  :

Cesium-137 8.87E-4 3.20E-4 ,

l A-3 1

I

TABLE A-2'(Continued)

-Quarter 1 Quarter 2 Barium-Lanthanum-140 1.80E-3 5.46E-3 Magnanese-54 1.06E-3 2.39E-4 Cobalt-58 1.80E-4 4.91E-5 Cobalt-60 3.39E-3 7.93E-4 Zinc-65 1.13E-4 3.16E-5 Zirconium-Niobium-95 5.27E-4 7.33E-5 Cerium-141 1.38E-4 1.03E-5 Cerium-144 3.95E-4 6.74E-5 Iron-59 1.95E-4 8.84E-6 Silver-110m 1.62E-4 NDA-Chromium-51 NDA 7.59E-6 GasesI "I Main Stack 5.41E 3 'l.39E 4 Reactor Bldg. Vent ,

5.99E 3 6.33E 3

• Unable to determine isotopes due to Augmented Off-Gas System f

NDA = No detectable activity V

i 1

g A-4 l

t: ,

i BOSTON EDISON COMPANY PILGRIM NUCLEAR GENERATING STATION Environmental Radiation Monitoring Program SEMIANNUAL REPORT NO. 9 JULY 1, ' 1976 THROUGH DECD1BER 31, 1976 I.

, .n. ,

w.

Prepared By i Joel 1. Cehn Environmental Sciences Group Nuclear Engineering Department

]

March 1977 4

Approved By:

G. Jame!(Davis, Group Leader Environt(ental Sciences

\ -

I. INTRODUCTION AND

SUMMARY

This report discusses the data accumulated by the Environmental'

~

Monitoring Program during the period July 1, through December 31, 1976.

Pilgrim Station reactor power output was about 75% of capacity during July, August and September.

The average capacity factor for October was about 70%,'and for November and December it was Labout 80%.

i

,. The Environmental Radiation Monitoring Program was performed during f( I the reporting period as. required by the Nuclear Regulatory Commis-sion. The most notable event was the detection of weapons test I fallout from the Chinese tests of September 26 and November 17.

Isotopes detected in' terrestrial media were I-131, Mn-54, Zr/Nb-95, Ba/La-140 and Ru-103, 106. I-131 was detected in milk at 51 pCi/1.

P Fallout nuclides detected in marine media included Mn-54, Zr/Nb-95, t

-Ru-103, La-140 and Ce-141, 144.

'. Plant related radioactivity was detected in marine algae, sediments and finfish. The highest level detected was 2.1+ .2 pCi Co-60 per gram Irish Moss.

l l

e l

1 j

... - m j

m c: -

=

TABLE 6 AIR PARTICULATES - GROSS BETA CONCENTRATIONS IN WEEKLY SAMPLES (pCi/m3 ) July - Dec 1976' East Plymouth Cleft Manomet Rocky Hill Property Overlook Pedestrian East Rocky Hill Period (1976) Weymouth Center Rock Area Substation Road (West) Line Area Bridge Breakwater Road (East) Warehouse July 1 - July 8 0.022 0.027 0.014 0.028 0.030 0.026 0.027 -0.029 0.027 0.029 0.022 0.028 0.031 0.027 0.032 0.028 0.031 0.027 0.030 0.030 0.023 0.031 July 8 - July 15 July 15 - July 22 0.025 0.034 0.023 C.024 0.031 0.026 0.022 0.031 0.024 0.020 0.028 July 22 - July 29 0.032 0.034 0.027 0.025 (d) 0.033 0.030 0.039 0.045 0.028 0.045 July 29 - Aug 5 0.031 0.033 0.026 0.028 0.020 0.027 0.027 0.024 0.028 0.027 0.034 Aug 5 - Aug 12 - 0.022 0.022 0.018 0.025 0.021 0.019 0.007(e) 0.021 0.018 0.030 0.027 Aug 12 - Aug 19 0.034 0.023 0.022 0.024 0.023 0.022 0.025 0.024 (d) 0.024 0.027 Aug 19 - Aug 26 0.037 0.036 0.038 0.034 0.039 0.039 0.032 0.035 0.030 0.043' O.038 Aug 26 - Sep 2 0.025 0.025 0.025, 0.028 0.031 0.041 0.027 0.039 0.034 0.034 'O.039

- Sep 2 - Sep 9 0.026 0.026 0.034 0.010(e) 0.023 0.026 0.018 0.030 0.025 0.021 0.018 Sep 9 - Sep 16 0.036 0.037 0.043 0.034 0.026 0.034 0.038 0.039 0.040 0.041 0.035-Sep 16 - Sep 23 0.040 0.034 0.027 0.031 0.039 0.032 0.033 0.04G 0.040 0.060 0.037 Sep 23 - Sep 30 0.039 0.044 0.031 0.037 0.037 0.033 0.030 0.038 (b) 0.028 0.043 Sep 30 - Oct 7 0.117' O.178 0.269 0.282 0.290 0.210 0.223 0.145 0.096(c) 0.168 0.181 Oct 7 - Oct 14 0.565 0.399 0.389 0.289 0.330 0.441 0.498 0.395 0.433 0.296' O.417 M Oct 14 - Oct 21 0.100 0.088 0.122 0.120 0.130 0.142 0.092 0.133 2.590 0.103 0.135

• Oct 21 - Oct 28 0.103 0.159 0.134 0.072 0.082 0.183 0.096 0.116 0.105 0.140 0.146 Oct 28 - Nov 4 0.202 0.346 0.260 0.278 0.391 (d) 0.262 0.302 0.231 0.250 0.297 +

Nov 4 - Nov 10 0.133 0.133 0.115 0.101 0.144 (d) 0.111 0.125 0.141 0.141 0.208 Nov 10 - Nov 18 0.195 0.180 0.135 0.143 0.178 0.084 0.163 0.161 0.162 0.140 0.134 Nov 18 - Nov 24 0.141 0.128 0.116 0.127 0.138 0.112 0.142 0.118 0.131 0.123 0.147 Nov 24 - Dec 2 0.145 0.137 0.094 0.140 0.12 (f) 0.100 0.13 (f) 0.129 0.08 (f) 0.14 (f) 0.10 (f)

Dec 2 - Dec 9 0.076 0.071 0.079 0.062 0.069 0.079 0.061 0.083 0.069 0.101 0.132 '

Dec 9 - Dec 16 0.096 0.132 0.094 0.090 0.101 0.102 0.098 0.093 0.103 'O.099 0.105 Dec 16 - Dec 23 0.048 0.078 0.047 0.073 0.068 0.063 0.071 0.081 0.071 0.064 (d)

Dec 23 - Dec 30 (b) 0.062 0.023 0.069 0.072 0.059 (b) 0.062 (b) 0.07 0.077 l

(a) Results of weekly beta counting of air particulate filters. Measurement error (2 e) is typically 0.008 pCi/m 3, ranging from 0.006 to 0.014.

(b) Station inaccessible (c) Two week sample (d) Station inoperable (e) Low sample volume - instrument malfunction ]

( (f) Average of two air samples taken sequentially during period ,

1

(

I I

.._.-.7._.- - ... . ..

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~s LTABLE:7~

AIR PARTICULATES - GROSS _ GAMMA CONCENTRATION.IN_ MONTHLY. COMPOSITES -

(cpm /m3 x 10-3).

i 4

Collection East Plymouth Cleft Manomet Rocky Hill. Property Overlook Pedestrian- East' Rocky Hill

- Period (1976) Weymouth Center- Rock Area Substation Road (West) Line Area Bridge Breakwater Road (East)' Warehouse.

July' 5+3 5+3 5+3 7+3 <4 6+3 7+3 9+3- _7+3 , 5+ 3 8+3 August 5+2 <2 2+2 7+2 9+2 5+2 <2 2+2 <3 3+2' 6+2 September <3 <6 <3 <3 <3 <3 <6- 613 <4; ' <3' <3 October 21+3 2113 12+3 17+3 17+3 19+3 15+3 1813 107+3 17+3 10+3. ,

November 2312 2113 24+2 16+2 20+3 32+4 2013 18+2 20+2 18 + 3 ~ 24+2 ,

'4+3 16+3(a) 17 3(a) -<3 20+4(a) 6+3 . 16+4(a)  !

December <4 <3 (3 (4 90 m

(a) No isotopes identified from gamma isotopic spectrum. ,

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TABLE 7B AIR PARTICULATES - STRONTIUM 89',90 CONCENTRATION IN QUARTERLY COMPOSITES (a) (pCi/m3 x 10-4}:

Collection East Plymouth Clefi Manomet Rocky Hill Property Overlook Pedestrian East Rocky Hill' Period (1976) Weymouth Center Rock Area, Substation Road (West) Line Area Bridge' Breakwater Road (East) Warehouse

~

Jan.-March Sr-89 (9 <25 <11 . <18 <12 <7- -15tl2 12 10 <11 <9 <ll

-Sr-90 2t1 <3 <1 3t1 4tl 2tl el 2t1 221 2t1- <1 April-June Sr-89 <8 <9 <10 6?6 (12 7t7 (10 <9 <8 '<9 <8.4

.Sr 3.621.5 St2 3t2 <11 2t2 2.7tl.4 4t2 412 5.911.5 2.7tl.5 1.921.3 July-Sept.

.s Sr-89 <20 <6 9t6 11*11 <12 <14 <9 <30 (20 <9.5 <9.1 Sr-90 <2 2.0+1.2 1.9*0.8 2.4+1.2 <l.5 3*2 2.6tl.2 323 312 3.6*1.1 3.021.1 Oct.-Dec.

Sr-89 <7 30920 80t30 60t20 220t20 90220 90t?0 100t30 100t20.. 90t30 Sr-90 1.610.8 <2 <5 <3 <2 2t2 212 <4 2*? <4 i

(a) Results of analyses per Regulatory Guide 4.6.

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

,_ . , ~ .

-- .l y 6 . . . .-

t TABLE 8 PARTICULATE IODINE-131 IN AIR SAMPLES (pCi/m3)

July - December Cleft Manomet li ky 11111 Property Overlook Pedestrian' East Rocky 11111 Collection East plymouth Period (1976) Weymouth Center Rock Area Substation m 4d (West) Line Area Bridge Breakwater Road (East) Warehouse

.0lt.02 .00*.02 .00t.02 .01t.02 .Olt.02 .002.02 .00t.02 .Olt.02 .011.02 Jul 1-Jul 8 .00t.02 .012.02

.00?.02 .00*.02 .0lt.02 .002.02 .00t.02 .00*.02 .00*.02 .0l*.02 .011.02 Jul 8-Jul 15 .002.02 .00t.02 .012.04

.02*.04 .Olt.04 .02t.03 .00*.04 .012.04 .024.03 .00t.04 .00t.04 .012.04 Jul 15-Jul 22 .00t.04 . cit.02

.01*.02 .00t.02 .00*.02 (d) .00?.02 .Olt.02 .001.02 .01*.021 .00?.02 Jul 22-Jul 29 .002.02 .002.02

.00t.02 .00*.02 .00t.02 .00t.02 .012.02 .012.02 .001.02 .Olt.02-Jul 29-Aug $ .Olt.02 .Olt.02

.01t.04 .0lt.04 .33t.04 .05t.07(e) .03t.04 .03*.04 .01t.04 .Olt.04 Aug 5-Aug.12 .02t.04 .02t.04 .01t.04

.0l*.02 .0l*.02 .001.02 .015*.014 .00t.02 .022.02 (d) .012.02 .012.02 Aug 12-Aug 19 .Olt.02 .00t.02 .012.03 Aug 19-Aug 26 .02t.03 .0l*.03 .01*.03 .01*.03 .00t.03 .00*.02 .Olt.03 .022.03 .011.03 .00t.02

.Olt.02 .0l?.02 .01*.02 .018t.015 .026t.014 .02t.02 .022t.014 .00t.02 .023t.014 .028t.015 Aug 26-Sep 2 .00t.02

.00*.04 .03t.05(e) .05t.02 .04*.02 .04?.03 .042.03 .042.02 .00t.03 .032.04 Sep 2-Sep 9 .02t.04 .03*.04

.02t.04 .04t.02 .02t.04 .06t.02 .002.04 .00*.04 .01t.04 Sep 9-Sep 16 +022.04 .04t.03 .02t.04 .0l*.03

.011.02 .011.02 Sep 16-Sep 23 .016t.014 .00*.02 .00t.02 .00t.02 .0159.014 .00*.02 .0lt.02 .012.02 .012.02 Sep 23-Sep 30 .01t.02 .012.02 .017t.013 .012.02 .0l*.02 .012.02 .01t.02 .018t.013 (b) .0l+ .0 2 - .012.02

.05*.02 .05*.02 .05t.02 .06t.02 .041.02 .018*.008(c) .04t.02 .06t.02 Sep 30-Oct 7 .Olt.02 .03*.02 .06t.02

.092.02 .09t.02 .06t.02 .081.02

$ Oct 7-Oct 14 .13t.02 .08t.02 .087.02 Oct 14-Oct 21 .018t.012.034?.013.019+.013 .035t.012

.08*.02 .08*.02

.032!.012

.122.02 .09t.02

.040t.012 .0349.013 .044t.012 0.02.02 (g) . 0 32 t .012 .0342.013

.Olt.02 .017*.014 .012.02 .00t.02 .018t.013 . 01 t . 02 .018t.014 .01f.02 .0231.015 .016t.014 Oct 21-Oct 28 .00t.02 (d) .0361.015 .046t.015 .03?.02 .05t.02 Oct 28-Nov 4 .027t.015.035t.015 039?.015 .031t.014 .051*.015 .03t.02

.00t.05 .05!.03 .02*.05 .04*.05 .05t.03 (d) .04*.05 .05t.03 .041.05 .05*.03 .06t.03 Nov 4-Nov 10 .019t.013

.025*.014 .01t.02 .017*.013 .0181.013 .0272.013 .Olt.02 Nov 10-Nov 18 .024t.013.022t.013.024*.013 .olt.02

.0l?.03 .012.03 .011.04 .041.02 Nov 18-Nov 24 .Olt.03 .03t.04 .03t.02 .04t.02 .02t.04 .012.03 .001.03 Nov 24-Dec 2 .017t.013 .01*.02 .016.02 .022t.012 .02t.08(f) .012.02 .Olt.08(f) .024t.013 .032.05(f) .02t.05(fl. .022.06(f)

.00?.02 .0l*.02 .004.02 .019.02 .02t.02 .00t.02 .01t.02 .00t.02 .00*.02 .021.02 Dec 2-Dec 9 .0lt.02

.0l?.02 .011.02 Dec 9-Dec 16 .00*.02 .00t.02 .00t.02 .022.02 .019.02 .011.02 .01t.02 .012.02 .002.02 Dec 16-Dec 23 .Olt.02 .00t.02 .012.02 .00t.02 .00t.02 .Olt.02 .012.02 .012.02 .012.02 .01t.02 (d)

Dec 23-Dec 30 (b) .019.04 .004.04 .02*.04 .02t.04 .02t.04 (b) .04t.03 (b) .01t.04 .041.03 ,

(a) Results of discriminated gamma counting (at 364 kev) of air particulate filters. Results are given in

• raw" form to permit averaging. Errors are t20.

(b) Station inaccessible.

(c) Two-week sample. 3 (d) Station inoperable.

(e) Low sample volume instrument malfunction.

(f) Average of two air samples taken sequentially during period.

(g) Calculated from Ge(Li) gamma spectrum.

i i

I i

l-_--._- - ____ _-

h z . ..

TABLE 8A (a)

GASEOUS IODINE-131 IN AIR SAMPLES (pCi/m3) JULY -' DECEMBER Control Offsite Onsite Collection East Plymouth Cleft Manomet Rocky 11111 Property Overlook Pedestrian East Rocky Hill Period (1976) Weymouth Center Rock Area Substation Road (West) Line Area Bridge Breakwater Road (East) Warehouse Jul 1-Jul 8 .Olt.04 .00t.04 .02+.04 .01*.04 .02+.04 .00t.04 .00t.04 .02t.04 .00t.04 .0l*.04 .00t.04 Jul 3-Jul 15 .00+.04 .Olt.04 .029.04 .0l+.04 .00+.04 .029.04 .Olt.04 .02t.04 .01t.04 .01*.04 .00t'04 Jul 15-Jul 22 .039.05 .022.05 .03t.05 .03*.05 .03t.05 .03t.05 .01+.04 .04 05 .04. 05 . 01 + . 0 4 . .0l+.05 Jul 22-Jul 29 .013 04 .02t.04 .02+.04 .00t.04 (d) .01. 04 .03 04 .05+.02 .03+.05 .00. 04 .00*.04 Jul 29-Aug 5 .03t.05 .03+.05 .02*.05 .02t.04 .02+.05 .0l+.05 .02*.05 .03t.05 .0 3 t .05 .02+.05 .02t.05 Aug 5-Aug 12 .03t.09 .02t.08 .12t.06 .0l+.07 .02s.08 .08+.05 .Olt.13(e) .072.08 .08t.06 .07*.08 .08t.05 Aug 12-Aug 19 .04t.03 .04t.05 .03t.04 .02t.04 .01t.04 .05t.03 .05t.03 .04t.03 (d) .04*.03 .04+.03' Aug 19-Aug 26 .03 05 .03t.05 .01t.05 .022.05 .03t.05 .Olt.05 .00t.05 .00t.05 .03+.05 .04t.05 .Olt.05 Aug 26-Sep 2 .01t.04 .021.04 .03t.04 .Olt.04 .039.04 .02+.04 .02+.05 .03t.04 .05+.03 .02*.04 .021.04 Sep 2-Sep 9 .02+.04 .00t.05 .03t.05 .02t.06(e) .00t.04 .03+.04 .02 05 .02t.05 .0l*.04 .03t.04 .03t.04 Sep 9-Sep 16 .05t.03 .03t.04 .05t.03 .0 3t.04 .04t.03 .03t.04 .03t.04 .03+.04 .04t.03 .05+.03 .04t.03 Sep 16-Sep 23 .0l*.04 .022.04 .02t.04 .00t.04 .01t.04 .00t.04 .01t.04 .02+.04 .00t.04 .02t.04 .021.04 Sep 23-Sep 30 .029.04 .011.04 .03t.04 .032.04 .0l+.04 .044.03 .03t.04 05+.03 (b) .02t.04 .03t.04 Sep 30-Oct 7 .10t.03 .02t.05 .57t.05 .10!.03 .26t.04 .09+.03 .10t.04 .08t.03 .02!.02(c) .12+.03 .06+.03 La C) Oct 7-Oct 14 .04t.03 .082.03 .13t.03 .Ilt.03 .07t.03 .llt.03 .10s.03 .05t.03 12 03 .08*.03 .06t.03 Oct 14-Oct 21 .03t.04 .07t.03 .12t.03 .07t.02 .04t.02 .07?.02 .03t.04 .06t.02 .09*.02(g) .08t.02 .04t.02 Oct 21-Oct 28 .02t.04 .03t.04 .06. 03 .02t.04 .00t.04 .01t.04 .001.04 .04t.03 .04+.05 .03t.04 .04t.03 Oct 28-Nov 4 .05t.03 .02*.04 .06+.03 .049.03 .02t.04 (d) .02t.04 .04t.03 .04*.03 .05t.03 .05t.03 Nov 4-Nov 10 .07t.10 .05t.10 .07t.10 .07+.10 .06t.10 (d) .04t.10 .05t.09 .09t.10 .109.07 .06?.09 Nov 10-Nov 18 .01t.04 .012.04 .02t.04 .00t.04 .02s.04 . 0 3 * . 0 5 . 01 t . 0 4 - .01t.04 .02t.04 .01t.04 .012.04 Nov 18-Nov 24 .00+.07 .01t.07 .00+.07 .Olt.07 .031.07 .03t.07 .05t.07 .04t.07 .06t.07 .023.07 .032.07(f)

Nov 24-Dec 2 .02s.04 .02t.04 .012 04 .03*.04 .05t.13(f) .0lt.04 .072.12(f) .00t.04 .03t.13(f) .07 13(f) .072.12 l

Dec 2-Dec 9 .01t.04 .02*.05 .00 05 .03*.05 .Olt.04 .00t.04 .02t.05 .02t.05 .022.03' .02t.05. .09t.03 Dec 9-Dec 16 .01t.05 .03t.04 .Olt.05 .03+.04 .Olt.05 .03+.04 .03+.05 .01t.05 .04t.05 .02 04 .02t.04 Dec 16-Dec 23 .04t.05 .01t.05 .02t.04 .02t.05 .02t.05 .00t.05 .02t.05 .00+.04 .04t.05 .02t.04 (d)

Dec 23-Dec 30 (b) .04t.09 .08t.09 .06t.09 .02t.09 .06t.09 (b) .04t.09 (b) .05t.09 .012.08 (a) Results of discriminated gamma counting (at 364 kev) of gaseous iodine filters. Results are given in " raw" form to permit averaging. Errors are 120.

(b) Station inaccessible.

(c) Two-week sample.

(d) Station inoperable.

(e) Low sample volume instrument malfunction.

(f) Average of two air samples taken sequentially during period.

.-,v . -- . , ,-

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

l l

i i

I I

TABLE IV-1 Cross Beta Activity in Air Particulates Observed Concentrations vs. Pgant Fredicted Contributions (a)

(pC1/s )

3rd Quarter (b) gg (c)

I g(d) gg(e)

Air-Particulate I 11(a) ~

$asolina Station Observed Predicted 11:1 Observed Predicted Its!

4 1.6x10*I 2.0x10-5 ~1.2x10*'

East Weymouth 3.1x10-2 2.4x10-5 7.7x10 (Background) 4 4 8.8x10*'

Flymouth Center 3.1x10*2 2.2x10 7.1x10*3 1.6x10*I 1.4x10 (Offsite) 2.7x10*2 4.5x10~3 0.17 1.4x10*I 3.5x10*3 0.02 Cleft Rock Area (Offsite)

~

5.9x10 ' O.02 1.4x10*I 9.7x10*3 0.07 Manomet substation 3.0x10-2

} (Offsite)

Rocky Hill Rd.(West) 2.9x10-2 1.2x10*3 0.04 1.6x10*I 8.0x10*' 5.0x10*3 (Onsite)

Property Line 3.0x10-2 5.7x10*' O.02 1.4x10*I 9.1x10*' 6.5x10~3

• (Onsite) 4 0.01 1.6x10*1 3.2x10~3 0.02 Overlook Area 2.8x10*2 4.0x10 m (Onsite)

Pedestrian Bridge 3.3x10-2 g,7,go-3 0.05 1.5x10*I 1.5x10-2 0.10 (Onsite) 3.1x10-2 1.5x10*3 0.05 3.5x10*III) 1.3x10-2 0.04 East Breakwater (Onsite) 4,3,to-4 0.01 1.4x10*I 1.5x10-2 o,gg Rocky Hill Rd.(East) 3.2x10-2 (Onsite) 3.3x10-2 8.2x10*' O.02 1.7x10*I 6.3x10~3 0.04 Warehouse (Onsite)

(a) observed gross beta concentrations in air are average quarterly values for each air particulate monitoring station (see Table 6). The gross beta concentrations in air predicted from plant ef fluent data and site specific meteorology are included for c ompa r ison.

(b) 3rd quarter extends from July 1 to September 30. 1976.

(c) 4th quarter extends from September 30 to December 30. 1976.

(d) column 1 indicates average quarterly stoss beta concentrations observed at each air sampling station (pC1/m3 )

(e) . column 11 indicates predicted gross beta concentration of air particulate due to plant effluents 20

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l, PILORIM NUCLEAR GENERATING STATION

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- Radioactive Effluent and Waste Disposal Report I including

[' Radiological Impact on Humans E

JANUARY 1 THROUGH JUNE 30,1977 r .,

.\

, s -

i l

l l

. . BY: NUCLEAR ENGINEERING DEPT. )

ENVIRONMENTAL SCIENCES GROUP l b.

DATE

'" SEPT.1,1977

'~

BOSTON EDISON COMPANY

)

,.J 1

. C

. 1

• e )

l .

l TABLE 1 A EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT ( . )

January through June 1977 '

GASEOUS EFFLUENTS-SUMMATIDN OF ALL RELEASES

)

Unis Quarter Quarter Est. Total 1 2 Enor, %

A. Fission & actintion gases

1. Totai release Ci 1.57BF5 1.82 45 3 .00 E+ .

2. Averaec release rate for period yCi/sec 20244 2 31 f4.

3. Percent of Technical specification limit  % 8 37 E40 g . 52 4d B. lodines

1. Total iodine.131 Ci 1 33 E-1 2 03 E-1 5 00 E 1

2. Average release rate for period pCi/sce 1 71 E.-2 2 58 L2 3 Percent of technical specification limit  % 2 415F0 34740 ,

C. Particulates

1. Particulates with half. lives >3 days ci 8.35E-2 7. 78 E-2 4 . 66 L+]

2. Average release rate for period uCi/see 1. 07 E-2 9.90 6-3

3. Percent of technical specifiestion hmit  % -

1 25E+C 5 93E 1

4. Gross alpha radioactivity Ci 2 66E-6 6 05E 7' D. Tritium

1. Total release C1 2.54541 2.12 Bel 6.42 R1

2. Average release rate for period yCi/sec 3 27 E+0 2 70 EF0

3. Percent of technicalspecification hmit  % N/A E N/A E

'3

TA8LE18

~ ~ ' ~ ~ ~ ~ ~ ~

' EFFlijENT AND WASTE DISPOSAL SEMIANNUAL REPORT .

January through June, 1977 GASEOUS EFFLUENTS-ELEVATED RELEASE .

CONTINUOUS MODE SATCH MODE Nuclides Released Unit Quarter Overter Quarter _ Ouarter

1. Fission gases krypton SS Ci 3.58 B&3 4.88 EF3 . E . E, krypton 45m Ci 1.83E+4 1.87SF4 . E . E krypton 47 Ci 1 12 F44 9 68 F43 . E . E krypton.88 Ci 3 60EF4 3 12EF4 . E . E xenon.133 Ci s . u EM 6 1 ? E54 E . E xen6n.135 Cl 1 49EF4 3 25 EF3 . E . E '

xenon.139m Ci 1 77 E4 .2 7. s7 E&2 . E . E xenon.138 Ci 9. n1 E51 7 10 Ep1 . E . E Others (specify) Ci E . E' E . E Ci 9 35EF2 9 96 EF2 E E vannn-133M .

Ci E . E . E . E unidentified Ci 9 69S&3 4 49SF4 E E Total for penod Ci 1 53 EF5 1 78Sp5 . E . E

2. todines lodine.131 Ci 5 83 & 2 9.60E2 . E . E iodine.133 Ci 1 08 El 2 56E-1 . E . -E iodine.135 Ci 2 17 E-1 2 84 F.f . E . E Total for period Ci 4.03 El 6.36S.1 . E . E

3. P$rticulates strontium.59 Ci 9 06 E-3 3 00 E-2 . E . E strontium.90 C 9 7 2 E.s 1 02IL4 . E . E cesium.134 Ci 8 15 E-6 1 01 E-5 . E . E cesium.137 C: 1 06 L3 6.10 64 . E . E banum4 ant hanum.140 Ci 3 57IL2 3 32 h-2 . E E Others (specoy) Ci . E . E . E . E mannanese - 54 Ca 2.59 65 9.04 66 . E E cobalt - 60 Ci 1 63 E-4 1 70 E-4 . E . E unidentitied .Ci . E . E . E . E i

cerium - 141 3.00E-5 1.15 E-5  ;

cerium - 144 5.09E-4 1.65 E-4 zirconium-N10BIUM-95 1.07E-6 N.D.A.

t TABLE 1C EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT i January-through June, 1977.

< GASEOUS EFFLUENTS-GROUND-LEVEL RELEASES CONTINUOUS MODE BATCH MODE Nuctides Released Urnt Quarter ' Quarter Quarter Quarter

1. Fision ;;ases krypton.85 Ci E' . E . E . E krypton S5m Ci 4,73E+0 NDA E . E E krypton.87 ~

Ci - 3. 291i+1 NDA E . E . E krypton 88 Ci 1.10E+1 NDA E . E E ,

xenon.133 Ci 1 09E+3 1. 79 E+2 . E . E xenon.135 Ci 2.13E+2 5.19 E+Z . E , E xenon.135m Ci 1 29E+3 2.95E+3 . E . E xenon.138 Ci 1 22E+1 wna E E E Others (specify) Ci . E . E . E . E Ci . E- . E . E . E Ci . E- . E . E . E unidentified Ci . E E E . E Total for period Ci 386 E+3 3.65E+3 . E .

El

2. lodines .

iodinc.131 Ci  ! 7 44 E-3 1.07E-1 . E' . E iodmc.133 Ci 2 44 E-1 2 18E-1 . E . E iodine 135 Ci 5 41 E-1 5 52E 1 . E . E__

Total for period Ci 8 .79 E-1 8. 77E.-1 . E . E

3. Particubtes strontium 49 Ci 3 86 E-3 5 46 E-3 . E . E st rontium.90 Ci 1. 64 E-3 1. 52 E-3 . E . E__

cesium.134 Ci 3. 86 E-4 5.46 6-4 . E . E cesium.137 C 1 08 E-3 9 015-4 . E . E barium lanttianum 140 C 2 94 E-2 5 60 L-3 . E . E

~

Otliers (specify) Ci . E , E . E . E mannanese - 54 Ci 4. 96 E-4 1. 4 2 E-4 . E . E cobalt - 60 Ci 6. 63 E-4 6. 86 E-4 . E . E unidentiGed Ci . E, . E . E . E cerium - 141 / .46r., -4 2. 55r.-5 cerium - 144 2.28E-4 NDA chromium - 51 NDA 7,60E-5

PILGRIM NUCLEAR POWER STATION RADIOACTIVE EFFLUENT AND WASTE DISPOSAL REPORT INCLUDING RADIOLOGICAL IMPACT ON HUMANS

~-

JULY l THRO'JGH DECEMBER 31, 1977 n _.

---v Prepared by:

fThomas

- ~~. Soudon L

Radiological Engineer Reviewed by:

/

/

, - - 9

• ct W" A-^ ^ * -

Susan L. Kannenberg

/

Senior Environmental Scientist Approved by: Al .

Fred JfMogolesid Manager of Environmental Science Group Date of Submittal: March 1, 1978

.- - .. . . . . - - ., . - . ~ . . - ___ -

i TABLE 1A-EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1977)

GASEOUS EFFLUENTS SUMMATION OF ALL RELEASES July - December,1977 Quarter Quarter . Est. Total Unit 3 4 Error, %

A. Fission and ac'tivation gases Ci 6.87E+ 4 5.30E+3 2.50E0 l

_ 1. Total release 8.64E+ 3 6.67E+ 2 2.' Average release rate for period uCi/sec

3. Percent of Technical Specification limit  % 3.51E0 3.75E-1 B. Iodines C1 1.44E-1 1.50E-3 2.81E+ 1 l

1. Total iodine 131

2. Average release rate for period yCi/sec 1.81E-2 1.89E-4

3. Percent of Technical Specification limit  % 2.28E0 1.13E-2 .

C. Particulates -

1. Particulates with half lives > 8 days Ci 3.87E-2 8.25E-3 2.84E+ 1 l

2. Average release rate for period pCi/sec 4.87E-3 1.04E-3

3. Percent of Technical Specification limit  % 2.93E-1 1.07E-1 l

4. Gross alpha radioactivity Ci 3.89E-7 < 1.14E-7 l

D. Tritium Ci 1.04E+ 1 3.71E0 3.40E+ 1 l

1. Total release

2. Average release rate for period pCi/sec 1.31E0 4.67E-1 l

3. Percent of Technical Specification limit  %

l 4

3 l

TABLEIB EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (YEAR)

GASEOUS EFFLUENT 9 - ELEVATED RELEASE July - Dec. enber,1977 CONTINUOUS MODE BATCH MODE Quarter Quarter l Ovarter l Unit l Quarter l l l Nuclides Released l

1. Fission gases krypton 85 Ci 2.36E0 3.07 E-3 Ci 8.22E+ 3 6.61E+ 2 k'rypton 85m Ci 6.12E+ 3 5.56E+ 2 krypton 87 -

Ci 1.61 E+ 4 1.46 E+ 3 krypton-88 Ci 2.90E+ 4 2.13 E+ 2 xenon 133 Ci 1.22 E+ 3 5.64 E+ 1 xenon 135 Ci 2.72E+ 2 8.46E+ 1 xenon-135m Ci 9.51E+ 2 2.82E+ 2 xenon 138 Ci 1.77E+ 3 7.21E+ 2 xenon 131m Ci 3.81 E+ 3 NDA xenon 137 Ci 4.75E+ 2 NDA xenon-133m Total for period Ci 6.79E+ 4 4.03E+ 3

2. Iodines Ci 7.45E-2 1.24 E-3 iodine-131 iodine 133 Ci 6.96E-2 1.06E-2 iodine 135 Ci 9.80E-2 175E-2 _

Total for period Ci 2.24 E-1 2.93E-2

3. Particulates strontium 89 Ci 9.84 E-3 3.06E-3 strontium 90 Ci 5.60E-5 2.25 E-5 cesium 134 Ci 5.42E-5 2.86E-5 cesium 137 Ci 2.65E-4 1.08E-4 barium lanthanum 140 Ci 2.13E-2 1.40E-3 chromium 51 Ci 2.61 E-5 NDA manganese 54 Ci 4.27 E-5 1.03 E-4 cobalt-58 Ci 1.78 E -6 2.33 E-6 fron.59 Ci NDA 6.79 E-6 cobalt-60 Ci 1.76 E-4 3.57 E-4 zine-65 Ci NDA 7.44E-6 zirconium niobium-95 Ci NDA 1.79E-5 cerium 141 Ci 2.08E-5 7.82E-6 cerium 144 Ci NDA 4.55 E-6 ruthenium 103 Ci 2.00E-6 1.76E-6 ruthenium 106 Ci 1.50E-5 NDA 4

TABLE 1C EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1977)

GASEOUS EFFLUENTS GROUND LEVEL RELEASE July - December,1977 CONTINUOUS MODE BATCH MODE Quarter Quarter Quarter Quarter Nuclides Released Unit -

1. Fission gases krypton 85 Ci NDA NDA krypton 85m Ci NDA NDA krypton-87 Ci NDA NDA krypton-88 Ci NDA NDA Ci 5.49E+ 2 9.18E+ 2 xenon 133 xenon 135 Ci 7.00E+1 1.17E+ 2 xenon-135m Ci 4.18E+ 1 . 6.99E+ 1 Ci 9.96E+ 1 1.67E+ 2 xenon 138 Total for period Ci 7.60E+ 2 1.27E+ 3

2. Iodines iodine-131 Ci 6.99E-2 2.60E-4 iodine-133 Ci 9.41E-2 1.06E-3 1.39E-3  !

lodine 135 Cl 1.44E-1 Total for period Ci 3.08E-1 2.71E-3

3. Particulates strontium 89 Ci 2.98E-3 2.98E-4 strontium 90 Ci 8.13E-6 5.67E-6 cesium 134 Ci 1.06E-4 9.60E-5 cesium.137 Ci 2.76E-4 2.12E-4 barium-lanthanum 140 Ci 2.18E-3 i 3.95E-4 manganese-54 Ci 1.84E-4 4.33E-4 cobalt-58 Ci 2.48E-5 1.52E-5 iron 59 Ci 1.89E-5 NDA cobalt-60 Ci 9.27 E-4 1.57E-3 zine-65 Ci 1.55E-5 1.06E-5 zirconium niobium 95 Ci 8.38E-6 9.47E-6 cerium 141 Ci 3.04 E-5 NDA ruthenium 103 Ci 5.83E-6 4.97E-6 ruthenium 106 Ci NDA 6.88E-5 e

5

]

4 F

' PILGRIM' NUCLEAR POWER STATION RADI0 ACTIVE EFFLUENT AND WASTE DISPOSAL REPORT INCLUDING RADIOLOGICAL IMPACT ON HUMANS _

JANUARY 1 THROUGH JUNE 30, 1978 Prepared by: h Thomas L. Sowdon Senior Radiological Engineer Approved by:

• Fred J. Mogolesko V i I

Manager of Environmental Science l Group l

I i

)

Date of Submittal: September 1, 1978

'l 2

TABLE 1 A EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1978)

GASEOUS EFFLUENTS SUMMATION OF ALL RELEASES January - June,1978 Quarter - Quarter Est. Total Unit 1 2 Error, %

A. Fission and activation gases Ci 1.67E+ 4 8.92E+ 3 2.75E+ 1 l

1. Total release yCi/sec 2.15E+ 3 1.13E+ 3

2. Average release rate for period

3. Percent of Technical Specification limit  % 8.96E-1 4.98E-1 B. Iodines Ci 1.96E-2 3.80E-2 3.91E+ 1 l

1. Totaliodine 131

2. Average release rate for period gCi/sec 2.57E-3 4.83E-3

3. Percent of Technical Specification limit  % 3.23E-1 3.92E-1 C. Particulates

1. Particulates with half-lives > 8 days Ci 1.77E-2 1.77E-2 3.75E+ 1 l

2. Average release rate for period yCi/see 2.28E-3 2.25E-3

3. Percent of Technical Specification limit  % 1.99E-1 1.81E-1

4. Gross alpha radioactivity Ci < 7.10E-7 < 2.72E-6 D. Tritium Ci 1.12E+ 1 2.06E+ 1 4.90E+ 1 I l

1. Total release

2. Average release rate for period uCi/sec 1.44E0 2.62E0

3. Percent of Technical Specification limit  %

l l

l l

1

-l4 I

3

TABLE 1B EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1978)

GASEOUS EFFLUENTS - ELEVATED RELEASE January June,1978 CONTINUOUS MODE BATCH MODE Unit Quarter Quarter l Quarter l Quarter l l Nuclides Released l l l

1. Fission gases krypton-85 Ci 4.98E-2 3.22E-2 krypton 85m Ci 2.83E+ 3 1.83E+ 3 krypton 87 Ci 2.66E+ 3 8.43E+ 2 krypton 88 Cl 6.77E+ 3 3.37 E+ 3 xenon.133 Ci 1.94 E+ 3 1.27E+ 3 xenon 135 Ci 1.18 E+ 3 1.21E+ 2 xenon.135m C1 2.24 E+ 2 1.79E+ 2 xenon 138 Ci 6.85E+ 2 7.31E+ 2 xenon 131m Ci xenon 137 Ci xenon 133m Ci Total for period Ci 1.63E+ 4 8.34E+ 3 l

2. Iodines iodine 131 Ci 9.91E-3 l 2.76E-2 i iodine 133 Ci 4.83E-2 l 1.49 E-1 l iodine-135 Cl 7.20 E-2 1.93E-1 Total for period Ci 1.30E-1 3.70E-1 1

3. Particulates l strontium 89 Ci l 3.69E-3 I 4.48E-3 l strontium 90 Ci l 2.73E-5 2.83 E-5 I cesium 134 Ci 6.63E-6 7.34E-7 cesium 137 Ci 1.16E-4 1.42E-4 l barium lanthanum 140 Ci 7.66E-3 7.52E-3 manganese 54 Ci 1.67 E-4 1.31E-4 cobalt 58 Ci 9.33E-7 cobalt-60 Ci 5.84E-4 2.92E-4 niobium 95 Ci 3.39 E-6 4.41 E-6 I cerium 141 Ci 9.91E-7 1.07 E-5 chromium 51 Ci 3.64 E-5 cerium 144 Ci 7.73E-5 ruthenium 106 Ci 1.29E-4 Ci .

Ci Ci i

9

TABLE 1C EFFLUENT AND WASTE DISPOSAL SEMI ANNUAL REPORT (1978)

GASEOUS EFFLUENTS GROUND LEVEL RELEASE January - June,1978 CONTINUOUS MODE BATCH MODE Quarter Quarter Quarter Quarter Nuclides Released Unit

1. Fission gases krypton 85 Ci krypton-85m Ci krypton 87 Ci krypton 88 Ci xenon 133 Ci Ci 4.45E+ 2 5.82E+ 2 xenon 135 xenon 135m Ci xenon-138 Ci Total for period Ci 4.45E+ 2 5.82E+ 2

! 2. Iodines iodine 131 Ci 9.73E-3 1.04E-2 iodine 133 Ci 7.21E-2 7.05E-2

~

iodine 135 Ci 1.37E-1 5.87E-2 Total for period Ci 2.19E-1 1.41E-1 f

i 3. Particulates strontium-89 Ci 9.28E-4 I 4.78E-4 4

strontium-90 Ci 7.05E-6 3.28E-6 cesium 134 Ci 7.03E-5 6.43E-5 cesium 137 Ci 3.52E-4 1.88E-4 barium lanthanum 140 Ci 1.69E-3 1.99E-3 chromium 51 Ci 4.49 E-5 1.97E-4 manganese-54 Ci 4.25E-4 2.91E-4 cobalt 58 Ci 6.53E-6 3.34E-5 1

cobalt 60 Ci 1.81E-3 1.40E-3 zine-65 Ci 1.12E-5 9.33E-6 zirconium-niobium 95 Ci 6.04E-6 1.94E-5 cerium 141 Ci 2.64E-5 3.05 E-5 cerium 144 Ci 1.15 E-5 7.54E-5 Ci .

5

PILGRIM NUCLEAR POWER' STATION

- RADI0 ACTIVE EFFLUENT AND WASTE. DISPOSAL. REPORT INCLUDING RADIOLOGICAL IMPACT ON HUMANS e

JULY l THROUGH DECEMBER 31, 1978  :

Prepared by: .

M cur *-  :[

- 4 4

We o r Thomas L. Sowdon Senior Radiological Engineer .

1 Approved by: . o o e Fred J' Mogoletko

. )

Group Leader of Environmental Sciences Group I

l March 1, 1979

~

Date of Submittal:  :

l l ,

i

r -s TABLE 1 A EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT.

GASEOUS EFFLUENTS. SUMMATION OF ALL RELEASES

' JULY - DECEMBER 1978 Quarter . Quarter Est. Total Unit 4 Error, %

3 A. Fission and activation gases

1. Total release Ci 4.24 E+3 2.83 E+3 5.00 E+1 l

2. Average release rate for period pCi/sec 5.33 E+2 3.56 E+2

3. Percent of Technical Specification limit  % 2.33 E-1 1.54 E-1 B. Iodines

1. Total iodine-131 Ci 2.66 E-2 4.00 E-2 3.75 E+1 l

2. Average release rate for period yCi/sec 3.35 E-3 5.03 E-3

3. Percent of Technical Specification limit  % 1.33 E0 2.00 EO C. Particulates

1. Particulates with half. lives > 8 days Ci 9.84 E-3 1.11 E-2 3.75 E+1 l

2. Average release rate for period pCi/see 1.24 E-3 1.40 E-3

3. Percent of Technical Specification limit  % 1.19 E-1 1.71 E-1

4. Gross alpha radioactivity Ci 4.4.43 E-7 (5.10 E-7 D. Tritium

1. Total release Ci 2.61 E+1 3.76 E+1 5.00 E+1 l

2. Average release rate for period pCi/sec 3.28 EO 4.73 EO

3. Percent of Technical Specification limit  %

• 1 3 l i

TABLE 18 EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT ( 1978)

GASEOUS EFFLUENTS - ELEVATED RELEASE JULY - DECEMBER, 1978 CONTINUOUS MODE BATCH MODE Quarter Quarter l Quarter l Quarter l l Nuclides Released l Unit l l

1. Fission gases krypton 85 - Ci 1.25 E-2 7.98 E-3 krypton 85m Ci 9.52 E+2 6.06 E+2 krypton 87 Ci 5.06 E+2 2.27 E+2 krypton-88 Ci 1.74 E+3 1.06 E+3 xenon 133 Ci 5.86 E+2 4.07 E+2 xenon-135 Ci 7.45 E+1 4.76 E+1 Ci 4.33 E+1 7.54 E+1 xenon 135m xenon-138 Ci 8.02 E+1 2.61 E+2 -

l Ci

Ci Ci Total for period Ci 3.98 E+3 2.68 E+3

2. Iodines iodine 131 Ci 1.69 E-2 2.44 E-2 lodine 133 Ci 6.47 E-2 8.08 E-2 iodine-135 Ci 7.53 E-2 6.00 E-2 Total for period Ci 1.57 E-1 1.65 E-1

3. Particulates strontium 89 Ci 1.40 E-3 1.45 E-3 strontium 90 Ci 2.83 E-5 1.20 E-5 cesium-134 Ci 2.30 E-6 2.49 E-6 cesium 137 Ci 3.31 E-5 4.04 E-5 barium lanthanum 140 Ci 4.86 E-3 4.22 E-3 chromium 51 Ci 1,49 p_5 manganese Ci 9.19 E-6 1.09 E-5 Ci Ci cobalt 60 Ci 3.27 E-5 4.11 E-s Ci Ci .

cerium 141 Ci 8.59 E-6 4.66 E-6 cerium 144 Ci 1.67 E-5 4.10 E-6 ruthenium 103 Ci 3.19 E-6 1.28 E-6 Ci 4.48 E-5 3.12 E-5 ruthenium 106 4

TABLE 1C EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1978)

GASEOUS EFFLUENTS GROUND LEVEL RELEASE JULY - DECEMBER 1978 CONTINUOUS MODE BATCH MODE Unit Quarter Quarter Quarter Quarter Nuclides Released l

1. Fission gases krypton 85 Ci krypton-85m Ci krypton-87 Ci krypton-88 Ci xenon 133 Ci s . 6 o v41 xenon 135 Ci 2.59 E+2 9.57 E+1 xenon 135m Ci ,

xenon.138 Ci Total for period Ci 2.59 E+2 1.51 E+2

2. Iodines iodine 131 Ci 9.67 E-3 1.56 E-2 lodine-133 Ci 7.08 E-2 1.14 E-1 iodine 135 Ci 1.32 E-1 1.99 E-1 Total for period Ci 2.12 E-1 3.29 E-1

3. Particulates strontium 89 Ci 6.75 E-4 8.25 E-4 strontium-90 Ci 3.30 E-6 4.96 E-6 cesium.134 Ci 2.38 E-s 2.os r-s cesiuta.137 Ci 9.59 E-5 1.06 E-4 barium lanthanum.140 Ci 2.36 E-3 3.99 E-3 manganese 54 Ci 1.58 E-5 1.17 E-5 cobalt-58 Ci 2.26 E-6 1.58 E-6 chromium-51 Ci 5.80 E-5 6.73 E-5 cobalt-60 Ci 1.33 E-4 8.86 E-5 zine-65 Ci 8.58 E-6 Ci cerium-141 Ci 2.45 E-5 8.85 E-5 cerium-144 Ci 9.46 E-6 Ci _ _

I 1

1 5

PILGRIM NUCLEAR POWER STATION RADI0 ACTIVE EFFLUENT AND WASTE DISPOSAL REPORT INCLUDING RADIOLOGICAL IMPACT ON HUMANS

^

(

JANUARY l THROUGH JUNE 30, 1979 4

4 Prepared by: -

Thomas L. Sowdon Senior Radiological Engineer Approved by

.

Fred J. Wogole5ko# ,

Group Leader of Environmental Science Group -

Date of Submittal: September 1, 197>

L TABLE 1 A EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1979)

GASEOUS EFFLUENTS SUMMATION OF ALL RELEASES l

Quarter Quarter Est. Total Unit 2 Error, %

1

' A. Fission and activation gases Ci 2.25E+3 3.33E+3 4.00E+1 l

1. Total release pCi/sec 2.89E+2 4.24E+2 j

2. Average release rate for period

3. Percent of Technical Specification limit  % 1.47E-1 2.01E-1 l B. Iodines f_ Ci 5.07E-2 2.61E-2 3.00E+1 l  !

l- 1. Totaliodine 131

2. Average release rate for period pCi/sec 6.52E-3 3.3?E-3~

l

3. Percent of Technical Specification limit  % 2.54E0 1.31E0 C. Particulates l

1. Particulates with half-lives > 8 days Ci 2.28E-2 41.17E-2 3.50E+1 l

2. Average release rate for period yC1/see 2.93E-3 e1.49E-3

3. Percent of Technical Specification limit  % 5.40E-1 e 2.13-1

4. Gross alpha radioactivity Ci 4 5.12E-7 4 4.25E-7 D. Tritium

1. Total release Ci 5.61E+1 1.84E+1 5.00E+1 l

2. Average release rate for period gCi/sec 7.22E0 2.34E0

3. Percent of Technical Specification limit  % NA NA 1

3

l TABLE 1B EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT ( 1979 )

GASEOUS EFFLUENTS - ELEVATED RELEASE CONTINUOUS MODE BATCH MODE Unit Quarter Quarter l Ouarter l Ouarter l l Nuclides Rekased l l l

1. Fission gases ,

krypton 85 C1 2.51E-2 1.82E-2 krypton 85m Ci 4.51E+2 5.01E+2 krypton 87 Ci 1.18E+1 7.30E+2 ~.

krypton-88 Ci 5.75E+2 1.19E+3 xenon 133 Ci 4.25E+2 3.00E+2 xenon.135 C1 2.66E+1 3.72E+1 xenon 135m Ci 7.06E+1 4.26E+1 xenon 138 Ci 2.84E+2 1.28E+2 xenon-131m Ci xenon-137 Ci xenon 133m Ci Total for period Ci 1.84E+3 2.93E+3

2. Iodines iodine 131 Ci 8.37E-3 2.14E-2 iodine 133 Ci 3.38E-2 2.88E-2 iodine-135 Ci 3.98E-2 2.84E-2 Total for period Cj 8.20E-2 7.86E-2

3. Particulates strontium-89 Ci 1.31E-3 1.58E-3 strontium 90 Ci 6.69E-6 8.82E-6 cesium 134 Ci 1.31E-6 7.13E-7 cesium 137 Ci 3.17E-5 4.20E-5 barium lanthanum-140 Ci 3.83E-3 3.25E-3 chromium-51 Ci manganese 54 Ci 4.85E-6 7.19E-6 cobalt 58 Ci 1.75E-6 iron 59 Ci cobalt-60 Ci 2.10E-5 2.11E-5 zine-65 Ci zirconium-niobium 95 Ci 7.54E-7 ,

cerium 141 Ci 2.08E-6 2.66E-6 cerium-144 Ci 5.32E-5 3.96E-5 ruthenium 103 Ci 1.84E-6 ruthenium 106 Ci 1.02E-4 1.58E-4 4

a . ..a ,,

l l

t TABLE 1C EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1979)

GASEOUS EFFLUENTS . GROUND LEVEL RELEASE CONTINUOUS MODE BATCH MODE Quarter Quarter Quarter Quarter j Nuclides Released Unit

1. Fission gases krypton 85 Ci 4 3.67E-5 44.10E-5 krypton 85m Cl 1.33E+1 1.48E+1 krypton 87 Ci 1.38E+2 3.86E+1 l krypton-88 Ci 6.61E+1 4.16E+1 xenon 133 Ci 2.85E+1 2.77E+1 1.61E+2 7.64E+1 l xenon 135 Ci Ci 3.27E+1 xenon 135m Ci 1.72E+2 i e xenon.138 Total for period Ci 4.07E+2 4.04E+2 l O

2. Iodines iodine-131 Ci 4.23E-2 4.69E-3 I lodine-133 Ci 3.05E-1 2.69E-2  !

iodine 135 Ci 6.19E-1 4.56E-2 i Total for period Ci 9.66E-1 7.72E-2 l

3. Particulates 1.47E-3 2.55E-3 i strontium 89 Ci strontium 90 Ci 1. 01 E-5 5.34E-6 I 4.89E-6 cesium 134 Ci 2.90E-6 cesium-137 C1 4.29E-5 7.85E-5 '

barium lanthanum 140 Cl 1.46E-2 e 3.87E-3 manganese 54 Ci 1.92E-5 8.59E-6 I cobalt 58 Ci 1

iron 59 Ci l cobalt-60 Ci 2.46E-4 7.96E-5 l l zinc 65 Ci zirconium niobium 95 Ci cerium 141 Ci 3.68E-4 4.92E-6 ruthenium 103 Ci ,

ruthenium 106 Ci l l 5

I

THE UNIVERSITY OFMAll Friday, October 23, 1992-i

~

David Mulligan F Commissioner.

Massachusetts Department of Public Health 150 Tremont Street I Boston, Massachusetts 02111 .)

Dear Commissioner Mulligan,

We respectively submit the report of the review committee for 1 the Southeastern Massachusetts Health Study of leukemia around the l Pilgrim power plant. The review committee met three times, held

)

one conference call, and have exchanged numerous faxed' copies of- l material and telephone calls. The conclusions reached by the l committee were arrived at by consensus. If you have any questions l j

please feel free to call any member of the committee.  !

Sincerely yours,

(- .

7 7 Jose L. Lyon, .D., .P.H.

Committee Co-Chairman l

Department of Family and Preventhe Medicine 50 North Medical Drive Salt Lake Cit >. Dah M132 (8:11 581 7234

--- _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ __