ML20153D891

Seismic Fragility of Big Rock Point Core Assembly & Reactor Vessel Supports

ML20153D891
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	10/31/1985
From:	Campbell R, Sues R, Warriner R STRUCTURAL MECHANICS ASSOCIATES
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ML20153D691	List: ML20153D687 ML20153D882 ML20153D891
References
SMA-13703.01, NUDOCS 8809060075
Download: ML20153D891 (35)

Text

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SMA 13703.01

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l SEISMIC FRAGILITY OF BIG ROCK POINT CORE ASSEMBLY, AND REACTOR VESSEL SUPPORTS I

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prepared for l

CONSUMERS POWER COMPANY Jackson, Michigan 1 October, 1985 I

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g g STRUCTURAL mECHAnlCS 080906o075 anonav

$R ADOCK 0500 1 5 ASSOCIATES WM ac...e...

5160 Bach Street. Newport Beach. Cef. 92660 (714) 833 7552

, I SMA 13703.01 4

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SEISMIC FRAGILITY OF BIG ROCK POINT CORE ASSEMBLY AND REACTOR VESSEL SUPPORTS 1

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I by R. D. Campbell R. W. Warriner R. Sues prepared for CONSUMERS POWER COMPANY Jackson, Michigan October,1985 I

I l g g STRUCTURAL mECHAnlCS

__ w- A S.S O C. eI .A.. Te,..E S 5160 Bach Street, Newport Beach, Cof. 92600 (714) 833 7552

s TABLE OF CONTENTS Section Title Page 1 INTRODUCTION . . . . . . . . . . . . . . . . . . 1-1 g 2 RESULTS AND CONCLUSIONS ............ 2-1 5 2.1 Reactor Scram .............. 2-1 2.2 Reactor Vessel Supports . . . . . . . . . . 2-2 3 DETERMINISTIC ANALYSIS OF THE REACTOR INTERNALS, CONTROL ROD DRIVE HOUSINGS, AND SUPPORTS . . . . 3-1 3.1 Reactor Internals and Control Rod 1 Drive System ............... 3-1 i 3.1.1 Analytical Models ......... 3-3 '

3.1.1.1 Control Rod Drive Housing . 3-3 3.1.1.2 Control Rod Drive Assembly. 3-4 3.1.1.3 Fuel Channel and Support 1 Tube ........... 3-4 3.1.1.4 Fuel Tube cuides ..... 3-5 3.1.1.5 Fuel Top Guide ...... 3-5 3.1.1.6 Core Support Plate .... 3-6 3.2 Reactor Vessel Support .......... 3-8 4 FRAGILITY ANALYSIS . . . . . . . . . . . . . . . 4-1 g 4.1 Core Support Plate ............ 4-1 I 4.1.1 Strength Factor .......... 4-1 4.1.2 Structural Response Factor . . . . . 4-3 4.1.3 Equipment Response Factor ..... 4-3 4.1.3.1 Peak Broadening and g Smoothing . . . . . . . . . 4-3 P 4.1.3.2 Damping . . . . . . . . . . 4-4 4.1.3.3 Modeling ......... 4-4 4.1.3.4 Mode Combination ..... 4-6 4.1.3.5 Earthquake Co:rponent Combi-nation .......... 4-6 4.1.3.6 Resultant Equipment Res-pense Factor ....... 4-7 i

TABLE OF CONTENTS (Continued)

Section Title Page 4.1.4 Core Support Plate Fragility . . . . . 4-7 4.2 Reactor Vessel Supports . . . . . . . . . . . 4-0 ,

4.2.1 Capacity Factor ........... 4-8 4.2.2 Structural Response Factor . . . . . . 4-10 ,

, 4.2.3 Equipment Response Factor ...... 4-11

( 4.2.4 RPV Support Fragility ........ 4-11 m REFERENCES

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1. INTRODUCTION This report was prepared as part of an integrated program to address outstanding issues identified in NUREG-0828, Integrated Plant Safety Assessment, Reference 1. The report addresses Topic III-6, Seismic Design Considerations. The Licensee has proposed to evaluate the seismic

' resistance of equipment important to safety using a combination of proba-bilistic methods and deterministic analysis.

This report addresses two technical issues, 1) the ability of the plant to SCRAM and 2) the integrity of the RPV supports. Simplified

[ deterministic analyses were conducted to establish a bisis for development of fragility curves for use in the Big Rock Point Probabilistic Risk Assessment Model.

A description of the simplified deterministic analyses and of L

the fragility development is presented. The fragility description is presented as a single fragility curvo defining conditional frequency of L failure versus peak ground acceleration at the site. Emphasis is placed on defining the minus one standard deviation capacity of the weakest link or links in the reactor internals and control rod drive units and the reactor pressure vessel support.

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2. RESULTS AND CONCLUSIONS 2.1 REACTOR SCRAM The ability of the reactor to scram during a seismic event is ,

governed by the capacity of the lower core support plate. Its lateral resistance is provided by four alignment brackets placed at 90 intervals.

During a seismic event, only two of the four brackets are effective in resisting lateral load. Each bracket is secured to the core support

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plate by two, one-half inch (1/2") diameter alignment pins and two one~nalf inch (1/2") diameter bolts. Lateral seismic loading on the core support plate places the pins and bolts in shear. Upon shearing of the pins and bolts, the core support plate can rmve laterally about one-half inch (1/2") until it uses up the clearance between the control rod drive housings and the penetrations in the core support plate.

" It is not evident that this shif t in the core assembly would

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preclude scrara. With the control rods completely withdrawn, a one-half

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inch misalignment would likely not keep the drive assemblies from starting to enter the core. But, during the insertion, the control rod blades would ultimately rub on the fuel assemblies and possibly bind. At the half insertion point, rubbing appears evident.

, The median ground acceleration capacity and the minus one loga-rithmic standard deviation were computed to be:

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v A = 0.369

-18 = 0.20g The core support plate is supported vertically in four locations

( aM must support the entire dead weight and vertical seismic reaction of the 80 fuel assemblies. Gross yielding of the core support plate in the 2-1

vertical direction does not result in significant misalignment of the control rod drives and the fuel assemblies and scram would likely not be precluded from this f ailure mode. The median ground acceleration capacity and the minus one logarithmic standard daviation were computed to be:

t X=0.409 l

-18 = 0.24g All other capacities of the reactor internals and control rod I drive assemblies are significantly higher. Their minus one logarithmic standard deviation capacity exceeds 0.39 peak ground acceleration.

2.2 REACTOR VESSEL. SUPPORTS The reactor pressure vessel (RPV) is supported by 24 two and one-half inch (2-1/2") diameter suppression rods constructed of SA 197-B7 alloy steel. The rods act as cantilever beams fixed at the reactor e

vessel od pinned at the reactor support structure. They are oriented a few deg ees off vertical so that vertical loading is carried in tension and lateral loading is carried in bending. There is a lower bumper support on the reactor vessel that provides lateral support to the lower portion of the RPV during an earthquake.

The fragility of the reactor supports was derived using low cycle f atigue theory and accounting for stress concentrations in the threaded rod end. Because of the large uncertainty in high strain, low cycle f atigue data and calculations, the minus one logarithmic standard deviation capacity is significantly below the median. The calculated values are:

X=0.659

-la = 0.219 1

L Details of the derivation are contained in Chapter 4, 2-2

3. DETERMINISTIC ANALYSIS OF THE REACTOR INTERNALS, CONTROL ROD DRIVE HOUSINGS, AND SUPPORTS 1

The reactor vessel, its internal structure and support system, is shown in Figure 3-1. In the event of a severe earthquake, the reactivity in the core must be terminated in order to achieve a stable shutdown. Control rods that terminate reactivity are inserted from the bottom of the reactor vessel into the core assembly. In the event of a reactor support failure, it is postulated that a large LOCA would occur.

3.1 REACTOR INTERNALS AND CONTROL ROD DRIVE SYSTEM There are two principal concerns with the ability to insert the control rods. -

1. Deformation of the control rod drive housings or the core assembly.

2. Crimping of the control rod drive discharge lines.

( Analysis was performed to determine the level of earthquake that would result in excessive deformation of the internals or CRD housings.

The threshold of excessive deformation was defined as either the onset of gross yielding of ductile components or the ultimate strength of brittle components.

The general approach was to first compute lg unit load response for all critical areas. Next, netural frequency estimates of components and assemblies were made using handbook formulae and the Rayleigh-Ritz method. Spectral accelerations were then estimated from the natural frequency results and the 19 Unit load responses were factored to corres-pond to the spectal acceleration resulting from a 0.12g Regulatory Guide Spectrum input at the structural base.

3-1

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3. DETERMINISTIC ANALYSIS OF THE REACTOR INTERNALS.

CONTROL ROD DRIVE HOUSINGS, AND SUPPORTS l

The reactor vessel, its internal structure and suoport system, is shown in Figure 3-1. In the event of a severe earthquake, the reactivity in the core must be terminated in order to achieve a stable shutdown. Control rods that terminate reactivity are inserted from the bottom of the reactor vessel into the core assembly. In the event of a I reactor support failure, it is postulated that a large LOCA would occur.

3.1 REACTOR INTERNAQ AND CONTROL ROD DRIVE SYSTEM There are two principal concerns with the ability to insert the control rods. -

l 1. Deformation of the control rod drive housings or the core assembly.

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2. Crimping of the control rod drive discharge lines.

Analysis was performed to determine the level of earthquake that would result in excessive deformation of the internals or CR0 housings.

The threshold of excessive deformation was defined as either the onset of gross yielding of ductile components or the ultimate strength of brittle components.

The general approach was to first c upute lg unit load response for all critical areas. Next, natural frequency estimates of components and assembiies were made using handbook formulae and the Rayleigh-Ritz method. Spectral accelerations were then estimated from the natural frequency results and the lg Unit load responses were factored to corres-pond to the spectal acceleration resulting from a 0.12g Regulatory Guide I Spectrum input at the structural base.

3-1 l

Reference 2 documents response results of the reactor building to a 0.12g Regulatory Guide Spectrum earthquake at the base of the containment. Floor spectra are provided at the location of the reactor vessel support but not for the reactor vessel; thus, specific spectra were not available for the reactor internals and control rod drive housing analyses.

The lateral fundamental frequency of the reactor vessel vibrating on its suspension rod supports was computed to be about 2.2 Hz without considering coupling and mass effects of the attached piping. These effects would tend to lower the fundamer.tal frequency of the vessel slightly. The vertical frequency of the vetal or its supports was I computed to be in the rigid range. Fundame'431 frequencies calculated for core components and assemblies were about three times the vessel horizontal frequency and it was postulated that t1e lateral ac eleration experienced by the rector internals and CRD housings would be aporoxi-mately equal to the rigid body response of the RPV. Considering that l

some higher frequency input could filter through the RPV supports, the reactor internals were considered to experience a median response. 25%

greater than the RPV rigid body response. Refinement of this assumption g would require the generation of spectra at the RPV from the coupled I structure and primary coolant loop model in Reference 2. Vertical input to the core assembly was taken as the vertical response spectrum St the RPV support since the supports were shown to be rigid in the vertical direction.

Crimping uf the control rod delve discharge piping was not considered a credible failure mode on the basis of analytical studies, experimental data and historical data. The discharge piping is one-half inch (1/2") diameter Schedule 80 austenitic stainless steel. Analytical studies conducted of piping systems subjected to seismic loading have demonstrated that piping systems can withstand large deformations without f ailure or complete collapse, References 3 and 4 Experiments co, ducted for the Electric Power Research Institute, Reference 5. have demonstrated 3-2 i

tha', ptring systems can easily withstand four times the allowable stress limit. .'or Leve O Service without fracture or excessive crimping.

g Reference 6 documents the performance of piping systems in earthquakes, y Pressure boundary ruptures have occurred in a few cases when the piping could not accomodate large deformations resulting from movement of equipment to which the piping was inchored. There are no known instances where sufficient deformation occurred that would crimp piping to a point that flow would be shut off. 1 It is therefore concluded on the basis of these recent'stui'es, that dynamic lcads will not crimp piping to the extent that flow restric-tions would preclude a scram.

3.1.1 Analytical Models Critical areas analyzed included:

a. Control rod drive housing

b. Control rod drive assembly I c. Fuel channel and support tube

d. Fuel tube guide j e. Fuel top guide assembly

f. Core support plate and alignmant brackets 3.1.1.1 Control Rod Drive Housing g The CR0 housing extending below the reactor vessel wat analyzed P as a cantilever beam. The full weight of the CR0 mechanism was distri-j buted over the length of this housing. A ig static load in both the f lateral and vertical directions resulted in low stresses in the housing and th? flange bolts. The fundamental frequency was computed to be greater than 10 Hz; thus, the response acceleration and resulting seismic stress is low for the 0.12g R.G. earthquake and the CR0 housings will not govero the scram fragility.

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3.1.1.2 Control Rod Drive Assembly The control rod drive assembly internal to the reactor vessel was evaluated for a lg lateral and vertical load. A hollow drive tube cantilevers out of the assembly at the level of the lower core support plate. During insertion of the cortrol rods, the tube is subjected to lateral loading due to its own inertia plus the lateral reaction from the control rcd blade assembly. A case where rods were half-inserted was considered to be a governing case. With rods fully withdrawn, the canti-lever length of the drive tube is essentially zero. We are not concerned about the case with the rods in the full-in position since, at this point, scram is already achieved. A half-insertion point was considered to be a reasonable location for evaluation. If gross yielding were to occur at this point, the scram is postulated to be incomplete. Effects of the water surrounding the tube and contained within the tube were included in the evaluation. For a lg latera~, and vertical load, the drive rod stress was computed to be about 7.6 ksi. The corresponding deflection was computed to be about 0.117 inches.

The fundamental frequency of the submerged rod in the half-in l position was computed to be about 8.6 Hz. This is well above the 2.2 Hz fundamental frequency of the vessel and the response to the 0.12g R.G.

spectrum is considerably less than the lg unit load case applied. Thus, the CR0 assembly was not considered a governing element.

3.1.1.3 Fuel Channel and Support Tube The fuel channil and support tube was modeled as a pinned beam subjected to lateral aN vertical acceleration. Reference 7 provides simplified methods to account for the dynamic coupling effects of closely-spaced square fuel assemblies in a body of water, l

i Maximum stress in the assembly occurs in the bolts that connect the support tube to the fuel channel. The stress was about 8.8 ksi for a l 19 unit load case.

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The fundamental frequency, including the coupling effects of the closely-spaced fuel assembly array and the water was about 5.7 Hz. This is sufficiently above the 2.2 Hz frequency of the Reactor Vessel that the fuel assembly will experience response less than lg for the 0.12g R.G.

spectrum input. The fuel assembly was determined not to be a governing weak link in the reactor internals.

3.1.1.4 Fuel Tube Guides The fuel support tubes are located and supported vertically and laterally by tube guidcs which are bolted to the lower core support plate.

The guides act as vertical cantilever beams, fixed at the core support plata and loaded at the support tube / tube guide interface. The tube guide attachment to the core support plate is via a one inch (1") diameter stud thro yh the core support plate secured by a nut on the underside of the plate. There is a one and three-quarter inch (1-3/4") diameter shoulder on the tube guide bearing on the top side of the core support plate.

Lateral loading from the fuel assembly is reacted in bending on the threaded connection.

Bending stress in the stud was compute be 14.4 ksi for a 19 lateral load on the fuel assembly. A plastic hin3a calculation was conducted accounting for the shift in neutral axis as the section goes plastic and it was determined that the capacity of the tube guide was 2.7 g lateral acceleration in the fuel assembly. The best estimate of the 0.12g R.G. spectrum acceleration in the fuel bundle was 0.62g resulting in a factor of safety to median hinge formation of 4.35. The elastic stress in the tube guide at 0.629 is 8.9 ksi. The tube guide is constructed of cast austenitic stainless steel material. The yield strength at 6000F is about 18.2 ksi, thus, the tube gu!de is substantially below yield for a 0.12g R.G. earthquake and is not the governing element.

3.1.1.5 Fuel Top Guidj _

The fuel enannel top guide is constructed of notched beams that ,

fit together in a rectangular grid wnrk. The top of the fuel channels 3-5

b fit within the rectangular grid. Vertical stop lugs are provided on the top guide beams to provide resistance to vertical loading. The top guide assembly hinges upward to allow removal of fuel bundles.

The beam assembly was examined for lateral and vertical loading.

During lateral loading, the center beams will react the greatest load.

The reaction of ten (10) fuel bundles is carried by each of the center beams in tension and is reacted at the periphery of the top guide assembly

{ through attachment bolts. Vertical upward motion results in bending of the top guide beams. Downward motion is carried by the lower core su9 port plate.  ;

There is no vertical loading on the beams until the effects of gravity are overcome. At about 1.39g total vertical acceleration, a full clastic hinge will form in the top guide beams. The vertical response to

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a 0.12g R.G. spectrum earthquake is about 0.44g; thus, the top guide capa-city is greater than three times the response from tnis reference earth-quake. The top guide beam is not the governing element.

3.1.1.6 Core Suppo,rt Plate The core 4? port plate provides vertical and lateral support to the lower end of the fuel assemblics. It is perforated to accomodate the control rod drives and to allow upward flow of the primary coolant.

Principal loading occurs from the 80 fuel assemblies.

The plate is supported vertically and horizontally at four locations along the outer diameter. Control rod drive penetration tubes enter the cutouts in the core support plate but, sufficient clearance exists between the tubes and plate holes that neither lateral or vertical support are afforded by the tubes.

The core support plate is too complex to analyze by hand. Conse-quently. a finite element model was constructed. The model shown in Figure 3-2 represents one-eighth of the support plate and is applicable

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h for vertical loading arising from the dead weight and vertical seismic reactions from the fuel assemblies. Plate bending elements in the ANSYS computer code were used in the model. Symetry boundaries were utilized to restrict the model size to a one-eignth (1/8) segment. A fairly coarse mesh size was used to limit modeling cost. The coarse mesh is justified on the basis that gross structural response is of prime interest and this can be captured by a coarse mesh.

The core support plate was found to be the most highly stressed ,

structural element in the core. The plate is one and one-half inches (1-1/2") thick and must support the 80 fu61 assemblies. The co abination of numerous cutouts and vertical support at only fcur points results in significant dead weight stress. Figure 3-2 shows the element and node numbering. Maximum dead weight stress occurs at Element 47 et the vertical support. The stress intensity at this point is 13.74 ksi due to dead weight. Near the center of the core Element 6 has a stress intensity of 12.73 ksi. Several elements in the plate have comparable stress intensities.

The core support plate is most sensitive to seismic loading in the vertical direction. Lateral acceleration results in some additional local asynnetric bending in the plate at fuel tube guide locations. The bending stress f v lateral load is less than 20% of that for vertical load for the same (1g) acceleration level.

Using the finite element results and scaling for vertical seismic response, the plate bending stress for the 0.12g R.G. spectrum plus dead weight is about 21.9 ksi. The ASME code would allow a primary bending stress intensity of 2.4 Sm or about 39.4 ksi for Level 0 Service loading, thus the plate meets the structural acceptance criteria of the

. code for the 0.12g R.G. earthquake.

The core support plate is positioned and supported laterally by four alignment brackets which attach to the plate and are supported by core support plate brackets which are fixed by welding to the reactor vessel, see Fig. 3-3. The core support plate brackets are slotted to allow radial thermal expansion of the support plate.. Lateral seismic 3-7

loading is carried by two of the four brackets. Each alignment bracket is secured to the core support plate by two one-half inch (1/2") diameter bolts and two one-half inch (1/2") diameter alignment pins. The bolts l and pins are austenitic stainless steel. They are subjected to shear l load resulting from lateral response of the fuel assemblies reacting through the core support plate. Eccentricity of the load outside the bolt and pin center results in amplification of the direct shear load.

i The ultimate shear capacity of the bolts and pins, using median material strength properties is 13.28 kips, whereas the lateral force from the 0.129 R.G. spectrum is aout 8.03 kips resulting in a median f actor of saf ety of 1.65 relative to the 0.129 R.G. earthquake. If the site-l specific earthquake (Reference 9) is considered, the f actor of safety is 2.60.

3.2 REACTOR VE,SEL Supp0RT g The reactor vessel is supported by 24 two and one-half inch B (2-1/2") diameter suspension rods about 30 inches long. The rods are oriented slightly off of vertical and carry the entire dead weight of the reactor vessel. The support rods are shown in Figure 3-1 and anchor to the support structure at about Elevation 610'. There are lateral bumpers at Elevation 599' which resist seismic motion. The reactor vessel center

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of gravity is near the elevation of the support rods and most of the lateral loading is carried by the rods.

The rods are free to swivel on the upper end but are fixed from rotation at the lower end by Jam nuts in each side of the reactor vessel support bracket. Lateral seismic loading is carried by cantilever bending of the support rods.

l The fundamental frequency of the reactor vessel was calculated to be about 2.2 Hz. For the 5% damped horizontal response spectrum at the RPV support, the spectral acceleration resulting from the 0.129 R.G.

earthquake is about 0.59 This results in a primary bending stress intensity ir, the rods of 147.5 ksi . The bolts are constructed of SA 197 i Grade B-7 material with a specified minimum yield strength of 105 ksi and 3-8 E

k ultimate strength of 125 ksi. Normal code allowable for Level 0 Service for a solid round section in bending is 0.75 Fy times a bump f actor for Level D Service. For the SA 193-87 material, the bump f actor computed per Appendix F of the code, F-1370, is 1.39. Thus, the code allowable stress would be 1.04 Sy. At a temperature of about 500 F, the allowable stress is about 92 ksi. If the site-specific spectrum were used in t.1e response calculaticn, the response and the resulting primary bending stress intensity would be 93.4 ksi, just slightly above the code allwable. Thus, the RPV supports would slightly exceed the intent of l current codes for the site-specific spectrum. There is still substantial i margin to failure since the suspension rods have barely exceeded yield.

l As discussed in Section 4.1.3.3, the analysis of the RPV was conducted by hand. The coupled RPV, primary coolant loop and reactor building model described in Reference 2 did not model the effects of the lwer lateral support, thus, an uncoupled hand calculation was conducted taking into account the effects of the lower lateral support. As discussed in Section 4.1.3.3, the results of the two analyses were close, thus, the coupling effects of the reactor coolant loop piping and the coupling to the reactor building and the effect of the lower lateral support were not significant. The uncertainty arising from these minor effects was taken into account in the fragility curve derivation.

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4. FRAGILITY ANALYSIS t

Fragility descriptions were developed for the governing mode of failure of the reactor internals (lower core support plate) and for the reactor vessel support. The methodology used for fragility development is the same as applied to other Big Rock Point Components, as documented in Reference 8.

4.1 CORE SUPPORT PLATE 4.1.1 Strength Factor The core support plate fragility is governed by the capacity of the alignment brackets to resist lateral load. There are four alignment brackets spaced at 900 intervals around the periphery of the core support plate. The brackets are slotted in the radial direction and fit over core support lugs welded to the reactc. vessel. The slotting allows free radial thermal expansion of the core support plate but restrains lateral movement.

During a seismic event, the equivalent of two of the four align-ment brackets will' carry lateral toading. The alignment brackets are attached to the core support plate by two, one-half inch (1/2") diameter bolts and two, one-half inch (1/2") diameter pins, both constructed of austenitic stainless steel. The bolts and pins are subjected to direct

( shear from the lateral response of the core support plate and sheer due to eccentricity of the loading relative to the centroid of the fastener pattern. The ultimate strength of each lug was calculated to be 13.28

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kips. The ultimate strength was based on median material strength of 1.25 times the code specified strength for stainhss steel. The code ultimate tensile strength is 59.3 ksi at 600 0F, Taking shear ultimate as 0.6 of tensile ultimate and representing enedia) strength as 1.25 times code, the ultimate shear strength was calculi.ced to be 44.5 ksi. For the geometry and cross-r.etions areas of the pins and bolts, this trans-lates to a loading captcity of 13.28 kips.

4-1

The calculated loading on each bracket was 8035 pounds for a 0.12g earthquake defined by a Regulatory Guide 1.60 spectrum. There are l no normk? lateral loads on the brackets. The strength factor of safety is then:

13.28 F

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=

g= . 65 There are two sources of uncertainty in this factor., the actual material strength and the failure mechamism.

Code material strengths are typicaM y set at about the 95% proba-bility of achievance level. Using the properties of the lognormal distri-bution, the uncertainty in material strength is:

=

8 3 h in(1.25)= 0.14 The median strength assumes that the ains and bolts fail simul-

[ taneously. There is zero clearance between the pins and their holes while the bolts have a clearance. In the worst case scenario, the pins

( would shear first before the bolts picked up load. Since the pin shear area is the greatest, the pin strength alone is a lower bound. The ratio of the pin area to total area of pins and bolts is 0.58, ':hus as a lower bound, the strength is only 58% of that calculated. Considering this a

-2 logarithmic standard deviation bound, the uncertainty is:

8 F = 1/2 t n (h) = 0.27 The combined uncertainty for capactty is then:

s c -(0.142 + 0.27')h = 0.31 4-2

J 4.1.2 Struct, ural Response Factor The structural response factor calculated in Reference 8 is applicable to the reactor vessel internals. It is based upon the site-specific spectrum anchored to 0.1089 being median, wherein the structural response analysis of Reference 2 is based on a 0.129 Regulatory Guide 1.60 spectrism. The resulting factor and uncertainty from Reference 8 is:

F RS

= 1.58 '

sRS

= 0.28 4.1.3 Equipment Response Factor There are several variables that contribute to uncertainty in the computed responses for the 0.12g R.G. spectrum. They include:

Peak broadening and smoothing Damping Modeling error Mode combination Earthquake component combination 4.1.3.1 Peak Broadening and Smoothing

/ Spectra provided in Reference 2 and used for computing response of the RPV have been peak broadened and smoothed. Raw spectra were not available, thus a generic factor and uncertainty are applied from other studies of comparing smoothed and beoadened spectra to raw spectra. The RPV fundamental frequency vibrating as a rigid body on its supports was computed to be 2.2 Hz. This is in a region where the spectral accelera-tion is rising rapidly with increased frequency. The real spectral accel-eration could be lowered by a substantial amount. Based on previous studies, a conservative estimate of the conservatism would result in a response factor of about 1.2. If unity is considered a -28 lower bound, the uncertainty is:

4-3

833 =1/2in(1.2)=0.09 where:

s F

33

= 1.2 l

4.1.3.2 Damping The reactor vessel response was based on 5% damping. This is considered median for a low frequency system whose main support elements are well into the plastic bending regime before failure. The damping ,

factor is then unity. Two percent damped spectral acceleration is I considered to be a lower bound on damping represented as minus 2  !

logarithmic standard deviation below the median. The uncertainty in the damping factor is calculated from.

S 8

0 = 1/2 in 3 2%

a%

5

= 1/2 in (h) = 0.15

( 4.1.3.3 Modeling The analysis of the reactor internals was very approximate and was conducted using idealized models and hand calculations. A sophisti-cated analysis would be based on a coupled dynamic model of the reactor building, the primary coolant loop, the reactor vessel and its internal components. Results reported in Reference 2 are based on a coupled model of the RPV, primary coolant loop and reactor building, however, the lower lateral support for the reactor vessel was not included in the model.

Spectra at the reactor building to RPV interface were provided in

( Reference ? but spectra on the RPV itself were not provided, thus, the response of the reactor vessel internals wss based on estimates of the acceleration that would be transmitted in'.c the vessel and to the internals.

There are three sources of uncertainty in the modeling. First, the effects of coupling must be considered. The uncoupled RPV frequency calculated by hand, considering the lower support effective and accounting for rotary inertia, was 2.19 Hz. As a simple check analysis, the entire 4-4

v.

mass of the RPV was assumed to be lumped on an equivalent spring repre-J senting the lateral suspension red stiffness and the resultant frequency was 2.11 Hz. Thus, the Reference 2 model which ignored the lower RPV s support was not a significant deviation from actual conditions. The L fundamental RPV frequencies from the Reference 2 coupled model were 1.92 and 1.99 Hz. Consequently, the response of the RPV at 2.2 Hz used in the

( fragility analysis is not particularly sensitive to the simple uncoupleri modeling assumptions. The RPV displacement at the suspension and supports was computed to be 1.02 inches from the 2.2 Hz uncoupled model and 1.21 inches maximum from the coupled model without the lower support. Assuming that the RPV internals response is directly proportional to RPV maximum displacement and assuming the difference in displacements of the two models to be a -28 spread, the uncertainty in coupling is calculated as:

8 g1 1/2 in (@)= 0.08 The second source of uncertainty stemt from the lack of having spectra applicable to the internals, i.e., spectra generated on the RDV.

The RPV responds as a rigid body at about 2.2 Hz. The fuel assembly fundau ntal frequency, including the co oling effects of closely-spaced rectangular fuel bundles in a body of water was computed to be 5.7 hz using the methods of Reference 7. Since the RPV response is predomi-nantly as a single rigid body mode ard the fundamental frequency of the internals is about 2.6 times as great, it is expected that the internals would respond as rigid bodies relative to the RPV. Accounting for the fact that some higher frequency vibrations can transmit through the RPV

( support system, the RPV internals were assumed to respond at 25% greater

acceleration than the RPV. At 2.2 Hz, the 5% damped spectral accelera-tion for the RPV was 0.5g for the 0.12g R.G. earthquake. The internals were assumed to experience 0.62g. Assuming 0.5g to be a minus one loga-rithmic standard deviation below the median, the uncertainty in internal response is calculated as

S M2 =tn( ) = 0.22 4-5 i.-m.-... . - - - -

The third source of uncertainty is in use of Reference 7 and idealized beam bending models to compute response of internals. The response is not particularly sensitive to frequency since the internals y are well .y from the frequency of the RPV vibrating on its support L system. But, the virtual mass effects of Reference 7 must contain a reasonable amount of uncertainty. This source of uncertainty was estimated to be represented by a logarithmic standard deviation of 0.2.

The combined uncertainty for modeling is then:

eg = (0.082 + 0.222 + 0.202 )b = 0.31 The modeling factor is assumed to be unity as there was no intended bias.

4.1.3.4 Mode Combination The internals were stated to respond essentially as rigid bodies relative to the RPV. There will be some higher frequency filtering through the RPV supports which could excite some higher modes of the in-

{

ternals. The effect is considered to be small and a mode combination uncertainty is assumed to be:

S = 0.1 MC The mode combination factor is considered to be unity.

4.1.3.5 Earthquake Component Combination The capacity was based on a single horizontal response direction wherein the response is governed by a vector of the two horizontal directions. The two horizontal responses at the RPV support are essen-

[ tially equal. Using the recomendations of Newmark, Reference 10, median response is reasonably represented by taking 100% of one component and 40%

{ of the orthogonal component in phase. If a 100% and 40% response are vector sumed, the resulting vector is 1.08 and the response factor is:

FECC = 14.08 = 0.93 4-6

j The absolute worst case of response would occur if the two orthogonal

, components were in phase. In that case, the vector sum would be 1.414 of the single direction response. Using 38 to represent an absolute bound, the uncertainty in the earthquake component combination is calculated as:

y 1 414 SECC = 1/3 in( g )= 0.09 4.1.3.6 Resultant Equipment Response Factor

( The equipment response factor is the product of the factors for each variable.

F =F 33 FD I F RE M MC FECC= 1.2(1)(1)(1)(0.93) = 1.12 The uncertainty is the square-root-of-the-sum-of-the-squares of the uncertainty in each factor:

B "

RE I8$S + 8b + 8k + 8kC + 8ECC)

=(0.09 + 0.15 + 0.31 + 0.1 + 0.09 ) = 0.38 4.1.4 Core Suoport Plate Fragility The median ground acceleration capacity is the product of the capacity factor, the structural response factor and the equipment response I factor times the peak ground accelerations of the reference earthquake.

d=1.65(1.58)(1.12)(0.12)=0.35g The uncertainty is the SRSS of the uncertainty for each of the factors.

8 C = (0.312 + 0.282 + 0.382 )b = 0.56

{

47

~ _ _ . .

e The minus one logarithmic standard deviation capacity is calcu-lated by:

[- -18 =dexp(-18)

L-

= 0.35 exp(-0.56) = 0.20 4.2 REACTOR VESSEL SUPPORTS i 4.2.1 Capacity Factor The reactor vessel support fragility was derived by a low cycle

{

fatigue analysis of the threaded suspension rods. The alternating stress due to the 0.12g R.G. spectrum earthquake, without stress concentration at the thread rods, was calculated to be 148 ksi. The ASME code states that the effective stress concentration factor need not exceed four.

( This is based on the net section shaking down to elastic action. This requirement is not met at the 0.12g R.G. earthquake level but would be

( very nearly met for the 0.119 site-specific earthquake spectrum. At this level of shaking, the average number of cycles to failure would be about 250 wherein a real earthquake would produce 3-5 strong motion cycles.

The strength factor must, however, be defined in terms of load rather than cycles to failure; therefore, a fatigue argument for high strain notched conditions must be developed. l 1

Reference 11 indicates that a theoretical stress concentration '

f actor, Kt , of 3.85 is appropriate for bolts with standard nuts.

However, the theoretical factor is generally conservative due to strain hardening and blunting in the notch. Actual notched fatigue data for similar alloys with ultimate tensile strength between 100 and 160 ksi were examined. For axial tests Figure 4-3 from Reference 12 shows the difference in the alternating stress capacity between unnotched and notched specimens with a Xt of 4.0. At 1000 cycles, the ratio is 1.5.

At 100 cycles, the ratio is only about 1.2. At the 100 cycle limit of the data in Figure 4-1, the unnotched specimen is well into the inelastic 4-8

L

[

range and the notched specimen is still elastic in the net section. This

[ convergence is because the tests were load controlled and by definition, notched and unnotched specimen data must converge at the ultimate tensile strength as long as thc material is not brittle. Bending data are more applicable to the problem at hand. Figure 4-2 shows bending data from rotating beam tests. More pronounced convergence is evident as the bending stress approaches the ultimate strength. Data from bending tests are not available in Reference 12 for Kt greater than 2.5, but, based upon the trends observed, it is concluded that an average fatigue life curve extr:cted from the ASME code design curve and a code-type fatigue analysis using a Kt of 4 is probably conservative at the 0.12g reference earthquake level. Beyond this level, the conservatism may disappear due to the gross plastic behavior in the net section, but the important reference point is the lower bound on failure. We know that there is a large margin at the 0.11g site-specific earthquake but are uncertain as

( to the median failure level.

For purposes of defining a fragility for the RPV suspension rods,

{

the ASME code fatigue methodology was applied. The ASME design fatigue curve has a factor of 20 on cycles built into the low cycle end. The slope of the curve is linear on a log-log plot at the low cycle end.

Figure 4-3 shows the ASME code design curve with an average curve super-( imposed. The data that form the basis for the design curve were gener-ated under strain-controlled conditions, thus the alternating stress

( scale is a convenience for relating linear elastic analysis to strain range.

Considering five cycles of strong motion to be the applied loading, the ASME code fatigue curve extrapolates to about 1.8x10 6 ,,

f alternating stress (about 6% strain). The applied alternating stress for the 0.129 R.G. earthquake, including a Kt of 4.0, is about 5.92x105 ksi and the strength factor is:

6 1.8x10 p

3

= 3.04 5.92x105 4-9

s The uncertainty on this value is quite large. The ASME code design curve

~

is a lower bound which accounts for data scatter, size effects, surface finish, etc. It does not have a defined confidence level but since all l'

~ the known data fall above the curve, it was assumed that the design curve E was gbout a 95 confidence value or 1.65 standard deviations below the average.

At 5 cycles, the extrapolated ASME code design curve allows an alternating stress of about 5.2x105 psi, Figure 4-3. The uncertainty in fatigue strength is then computed to be:

6 s

3 =1/1.65in(1.8x10) = 0.75 l 5.2x105 l

( There is also uncertainty in using linear elastic fatigue criteria beyond the point where the net section remains elastic. The net

( section is elastic up to about the 0.119 site-specific earthquake level.

Beyond that level, strain concentrations at the thread roots may increase.

It was assumed that the uncertainty due to extrapolating is equal to the uncertainty in the a'terage fatigue curve, i.e.,

( B U = 0.75

( The combined uncertainty on the fatigue strength is then:

B =(0.752 + 0.752 )b = 1.06

[ C These uncertainty assumptions are benchmarked against a deterministic design fatigue calculation in Section 4.2.4.

( 4.2.2 Structural Response Factor The structural response factor developed for the RPV internals is applicable.

F g3 = 1.58 o g3 = 0.28 4-10

4.2.3 Equipment Response Factor The variables that make up the equipment response factor are the same as those described for the RPV internals. The factors and uncertain-

- ties are likewise identical, except for tne modeling factor. Since the L RPV responds as a rigid body on its supports and the coupling effects between structure, primary coolant piping and RPV were shown to be small,

( the uncertainty in modeling is judged to be equal to the RPV/ structural piping coupling effects described for the RPV internals.

sg = 0.08

[ The modeling factor is considered unity.

( The overall equipment response factor and uncertainty for the RPV supports is:

( F

• l'l2 RE s = 0.24 RE

( 4.2.4 RPV Support Fragility The RPV support fragility is the product of the capacity, struc-

{ tural response and equipment response factors times the reference 0.12g R.G. earthquake.

d=F3Fg3 FRE (0 12) = 3.04(1.58)(1.12)(0.12) = 0.65g The uncertainty is the SRSS of the uncertainties for the three factors.

= +

8 C ' (8k

• 8kS

• 8RE) . .8 + 0.2C h = 1.12 The minus one logarithmic standard deviation value of capacity is:

-18 = 0.65 exp(-1.12) = 0.21g 4 11 9

J As a check on the reasonableness of the analytical model, the 95%

~

confidence value of peak ground acceleration capacity is compared to the code design fatigue requirements for a regulatory guide spectrum. Using J a code Kt of 4.0 and the 0.12 regulatory guide spectrum response, the L alternating stress in the suspension rods was calculated to be about 592 ksi. From Figure 4-3, the design allowable alternating stress extrapo-( lated to 5 cycles is 520 ksi, thus the allowable acceleration level would be:

A,)jo, = h (0.12) = 0.106 The 95% confidence acceleration from the fragility model is:

[ A 95

= 0.65 exp(-1.65)(1.12) = 0.102 ,

( These values compare favorably and it is concluded that the lower tail of the fragility curve is reasonable.

(

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80 3 X N r UTS 115-130 KSI times design number of cycles.

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S W

5. REFERENCES L 1. NUREG-0828, "Integrated Plant Safety Assessment, Systematic '

l Evaluation Program", Consumers Power Company, Docket No. 50-155, September, 1983.

2. D'Appolonia Report 78-435, "Seismic Safety Margin Evaluation, Reactor Building Primary Coolant Loop, Volume II, Appendices A and B", September, 1980.

{

3. Campbell, R. D., R. P. Kennedy and R. D. Thrasher, "Development of l Oynamic Stress Criteria for Design of Nuclear Piping Systems",

( Structural Mechanics Associates Report, SMA 17401.01, prepared for the Pressure Vessel Research Committee, March, 1983,,

4. Broman, R., A. P. Cimento, R. Gamble, O. Leong and K. C. Warapius,

[- "Conceptual Task to Develop Revised Dynamic Code Criteria for Piping", EDS Nuclear Draft Report, prepared for Electric Power Research Institute, March, 1983.

I

5. ANCO Engineers Report 1182.13, "Laboratory Studies: Dynamic Response of Prototypical Pipire; Systems", June,1984.

6. NUREG-1061 Volume 2 Addendum, "Summary and Evaluation of Historical Strong Motion Earthquake Seismic Response and Damage to Aboveground Industrial Piping", prepared by Stevenson and Associates.

7. Shimogo, T. and Y. Shinohara, "Vibrations of Square and Hexagonal Cylinders in a Liquid", ASME Paper 80-C2/PVP-97.

8. Wesley, D. A., R. O. Campbell and R. Peek, "Seismic Capacities of Selected Big Rock Point Structures and Components", SMA Report 13703.01R002, April, 1933.

9. NRC Letter, D. Crutchfield to SEP Owners Group, June 8,1981.

10. Newmark, N. M. and W. J. Hall, "Development of Criteria for Seismic Review of Selected Nuclear Power Plants" NUREG/CR-0098, May, 1978.

11. Peterson, R. E., Stress Concentration Design Factors, John Wiley and Sons, 1953.

12. NA55-866, Fatigue Manual, North American Aviation, Rev. 5-72-64.

9 4

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[ EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT FOR FIRST AND SECOND QUARTERS, 1988

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Yankee Atomic Electric Company Rowe, Massachusetts 4824R/20.124 I b lb 4 I 8809060329 000630 ADOCK 0000 9 gDR

L TABLE 1A Yankee Atomic Electric Company. Rowe Massachusetts Effluent and Waste Disposal Semiannual Report First and Second Quarters. 1988

{-

Gaseous Ef fluents - Sununation of All Releases Unit Quarter Quarter Est. Total

( 1 2 Error. %

A. Fission and Activation Gases

[ 1. Total release C1 5.5SE+01 5.79E+01 z5.50E+01

2. Averate release rate for period uCi/sec 7.10E+00 7.37E+00

3. Percent of Tech. Spec. limit (1)(4)  % 5.40E-01 5.96E-01 B. Iodines

1. Total Iodine-131 Ci 1.12E-06 3.28E-06 22.50E+01

2. Average release rate for period uCi/see 1.42E-07 4.17E-07

3. Percent of Tech. Spec. limit (2)(4) 4.01E-01 5.67E-01

{  %

C. Particulates

1. Particulates with T-1/2 > 8 days C1 1.56E-06 1.93E-06 23.00E+01

2. Average release rate for period uCi/see 1.98E-07 2.46E-07 (3) (3)

( 3.

4.

Percent of Tech. Spec. limit Gross alpha radioactivity Ci <5.40E-08 2.48E-08

[ D. Tritium p 1. Tota.1 release C1 1.37E+00 1.18E+00 23.00E+01 L 2. Averare_ release rate for period uCi/see 1.74E-01 1.50E-01

3. Percent of Tech. Spec limit 1 (3) (3)

(1) Technical Specification 3.11.2.2.a for sanna air dose. Percent values for Technical Specification 3.11.2.2.b for beta air dose are approximately the same.

(2) Technical Specification 3.11.2.3.a for dose from I-131. Tritium, and radionuclides in particulate form.

(3) Per Technical Specification 3.11.2.3, dose contributions from Tritium and particulates are included with I-131 above in Part B.

(4) The first and second quarter percent of Technical Specification limits are baced on conservative plant quarterly dose dete'rminations.

4824R/20.124 f - - - - - - - - - - - --- -

TABLE IB Yankee Atemic Electric Company. Rowe, Massachusetts e

Effluent and Waste Disposal Semiannual Report First and Second Quarters, 1988 Gaseous Effluents - Elevated Release Continuous Mode Batch Mode (1)

Nuclides Released Unit Quarter Quarter Quarter Quarter I 1 2

1. Fission Cases I Krypton-85 Krypton-85m Ci Ci 2.96E-02 5.59E-01 3.74E-02 6.61E-01 Krypton-87 Ci 5.13E-01 5.58E-01 Ci 1.02E+00 1.22E+00 I Krypton-88 Xenon-133 Xenon-135 Ci Ci 2.47E+01 1.16E+01 2.36E+01 1.30E+01 ___

d Xenon-135m Ci 1.61E+01 1.68E+01 I Xenon-138 Xenon-133m Ci Ci 1.48E-01 4.89E-01 3.75E-01 7.25E-01 Argon-37 Ci 4.92E-02 6.01E-02 Argon-41 Ci 2.02E-01 2.68E-01 1 Carbon-14 Ci 6.10E-03 7.71E-03 Xenon-131m Ci 3.91E-01 5.80E-01 i Unidentified Ci - -

_ Total for period Ci 5.58E+01 5.79E+01

2. Iodires Iodine-131 Ci 1.12E-06 3.28E-06 1 Iodine-133 Ci 6.30E-07 2. 58 E- 06 Iouine-135 Ci <1.40E-07 (5.70E-07 Total for period Ci 1.75E-06 5.86E-06

3. Particulates Strontium-89 Ci (7.77E-08 <2.31E-07 i Strontium _90 Cesium-134 C1 Ci

<1.01E-07 2.61E-08

<l.85E-07 (4.32E-07 Ce41um-137 Ci 4.13 E-0 7 3.13E-07 Ba r ium-Lan t hanum-140 Ci <9.23E-07 <1.29E-06 Zine-05 Ci (7.18E-07 <1.01E-06 Cobalt-58 Ci (3.02E-07 <4.42E-07 Cobalt-60 Ci 1.09E-06 1.62E-06 I Iron-59 Chromium-51 Ci Ci

<6.76E-07

<l.982-06 (9.22E-07

<2.54E-06 Zirconin.n-Niobium 95 Ci (4.99E-07 <7.33E-07 I Cerium-141 Cerium-144 Ci Ci

<2.(0E-07

<1.14E-06

<3.56E-07

<l.58E-06_

Antimony-124 Ci (2.84E-07 <3.90E-07 Manganese-54 Ci <3.24E-07 ( 4. 5 3 d'-0 7 1 Silver-110M Ci <2.84E-07 <3.79E-07 Molybdenum-99 Ci <2.12E-06 (2.98E-06 Ruthenium-103 Ci < 2. 5_8 E .0 7 <3.48E-07 Total for period Ci 1.53E-06 1.93E-06 1

(1) T5are were no batch mode releases during this reporting period.

Abl 4t<D 124

TABLE 1C Yankee Atomic Electric Company, Rowe, Massachusett_s Effluent and Waste Disposal Semiannua1 Repot.

" First and Secend Quarters, 1988 Caseous Eff1uents - Ground Level Releases i There were no routine measured ground level continuous or batch mode gaseous releases during the first or second quarters of 1988.

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l TABLE 2A Yankee Atomic Electric Company. Rowe. etassachusetts F.ffluent and Waste Disposal Semiannual Report First and Second Quarters. 1988 Liquid Effluents - Summation of All Re; eases Unit Quarter Quarter Est. Total 1 2 Error. %

A. Fission and Activation Products

1. Total release (not including tritium, gases, alpha) Ci 5.96E-03 2.59E-02 22.00E+01

2. Average diluted concentration 1 during period uCi/ml 9.67E-11 4.20E.10

3. Percent of applicable li.46 (1)  % 2 . 8 9 E-0_4, 1.57E-03 B. Tritium

1. Total release Ci 4.45E+01 6.69E+01 21.00E+01__

i 2. Average diluted concentration during period uC1/mi 7.22E-07 1.08E-06

3. Percent of applicable limit (1)  % 2.41E-02 3.61E-02 C .fDissolved and Entrained Cases

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1. Total release Ci 1.40E-02 3.09E-01 22.00E+01 i 2. Average diluted concentration during period uC_i /ml 2.27E-10 5.01E-09

3. Percent of applicable limit (2)  % 1.14E-04 2.50E-03 D. Cross Alpha Radioactivity

1. Total release Ci <2.04E-06 <3.97E-06 23.50T+01 E. Volume of liquid effluent released (prior to dilution) liters 5.26E+06 7.14E+06 23.00E+01 1

F. Volume of dilution water used during period liters 6.16E+10 6.17E+10 25.00E+00 (1) Concentration limits specified in 10CFR, Part 20, Appendix B. Table II, Column 2 (Technical Specification 3.11.1.1). The percent of applicable limit reported is based on the average diluted concentration during the period. At no time did any release exceed the concentration limit.

(2) Concentration limits for dissolved and entrained noble gases ir 2E-04 microcuries/ml (Technical Specification 3.11.1). The percent i. applicable limit reported is based on the average diluted concentration during the period. At no time did any release exceed the concentration limit.

4824R/20.124

l TABLE 2B Yankee Atomic Electric Company, Rowe, Massachusetts Effluent and Waste Disposal Semiannual Report ,

First and Second Quarters, 1988 Liquid Effluents Continuous Mode Batch Mode I _, Nuclides Released Unit Quarter 1

Quarter 2

Quarter 1

Quarter 2

1 Strontium-89 Ci <1.35E-04 <2.39E-04 (5 59E-06 <1.88E-05 Strontium-90 Ci (1.55E-04 <1.13E-04 <8.02E-06 <9.14E-06 Cesium-134 Ci 3.55E-06 1.47E-05 1.24E-04 1.08E-03 I Cesium-137 Ci Ci 6.66E-06 4.05E-06 4.21E-05 9.10E-05 1.72E-04 3.95E-05 1.22E-03 1.25E-04 lodine-131 I Coba!t-58 Ci <2.74E-05 <2.38E-05 <3.84E-06 (7.74E-06 Cobalt-60 Ci 6.14E-06 5.02E-06 1.76E-05 3.15E-05 (4.98E-05 (7.43E-06 <1.32E-05 I Iran-59 Zine-65 Ci Ci

<5.91E-05 (6.14E-05 (5.20E-05 <8.59E-06 <1.51E-05 2.88E-07 Hanganese-54 Ci <2.75E-05 <2.38E-05 <4.12E-06 Chromium-51 Ci (2.01E-G. <1.81E-04 (4.60E-05 <1.22E-04 Zi rconimt -Niob ium-95 Ci (4.71E-05 (4.01E-05 ( 6 . 9_4 E-0 6 <1.51E-05

<1.76E-04 <3.00E-05 I Molybdenum-99 Technetium-99m Ci Ci

<1.94E-04 4.78E-06 1.32E-06 4.67E-07 (6.62E-05_

<1.65E-05 Bariam-1.an Wanum-140 Ci <8.97E-05 (7.91E-05 <2.04E-05 (5.18E-05 Cerium-141 Ci <3.44E-05 <3.13E-05 <7.76E-06 r2.65E-05 Ruthenium-103 Ci <2.63E-05 <2.25E-05 <5.63E-06 <1.52E-05 I Cerium-144 Todine-135 Ci Ci (1.54E-04

<2.57E-05 (1.39E-04 3.12E-05

<3.55E-05 1.61E-06

<1.21E-04 2.19E-06 Selenium-75 Ci <2.77E-05 <2.47E-05 <6.33E-06 < 1. 7 ?.E-0 5 Silver-110m Ci (2.57E-05 <2.23E-05 <5.38E-06 <9.66E-06 Antimony-124 Ci (2.48E-05 <2.28E-05 (7.86E-06 <2.05E-05_

carbon-14 Ci - - 5.59E-03 2.26E-02 Ci <7.01E-04 (5.78E-04 <2.92E-05 6.10E-04 l Iron-55 I Cesium _-136 Antimony-125 Ci Ci

<2.67E-05

<7.10E-05

<2.36E-05

<6.26E-05

<3.97E-06

<1.87E-05 3.18E-06

<4.77E-05 Unidentified Ci - - - -

I Total for period (above) Ci 2.52E-05 1.85E-04 5.94E-03 2.57E-02

_ Xenon-133 Ci 5.49E-05 2.19E-06 1.26E-02 2.91E-01 Xenon-135 Ci <2.18E-05 <1.84E-05 4.64E-05 1.45E-04 I Xenon-i31m Xenon-133m Ci Ci

<8.65E-04

<1.74E-04 (7.95E-04

<1.54E-04 2.06E-04 7.87E-05 5.21E-03 1.48E-03 Krypton-85 Ci (8.87E-03 (7.91E-03 1.00E-03 1.1/E-02

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TABLE 3 Yankee Atomic F.lectric Company, Rowe, Massachusetts Effluent and Waste Disposal Semlannual Report First and Second Quarters, 1988 Solid Waste and Irradiated Fuel Shipments A. Solid Waste Shipped Off-Site for Burial or Disposal (Not Irradiated Fuel)*

IJnit Six-Month Est. Total Period Error, %

1. Type of Waste

a. Spent resins, filter sludges, evaporator m3 bottoms, etc. - LSA container **,+ Ci

b. Dry compressible waste, contaminated mJ 3.27E+01 i equipment, etc. - LSA container ++

c. Irradiated components, control rods, etc.

Cit m3 Ci 2.34E-01 21.00E+02

d. 23 I e.

C1 mJ Ci 1 2. Estimate of Major Nuclide Composition (By Type of Waste)***

b. Cesium-137  % 3.23E+01  %

I Cesium-134  % 2.97E+01  %

Iron-55  % 2.65E+01  %

Cobalt-60 1 3.93E+00  %

i Niobiura-95 Nickel-63

% 2.69E+00

% 2.06E+00 Iron-59  % 1.58E+00  %

Manganese-54  % 1.10E+00  %

I Carbon-14  % 4.70E-02 _

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3. Solid Waste Disposition Number of Shipments Mode of Transportation Destination 1 Truck Barnwell, SC I B. Irradiated Fuel Shipments (Disposition): None

+ Container volume equal to 5' gallons (drums).

• Container volume equal to 105 ft3 (boxes).

• Solid waste is Class A, as defined in 10CFR61.55.

• • Solidification agent is cement.

• • • Excluding nuclidos with half-lives less than 12.8 days.

t Estimated.

4824R/20.124

L APPENDfX A

{ Radioactive Liquid Effluent Monitoring Instrumentation Requirement: Radioactive liquid effluent monitoring instrumentation channels

{ are required to be operable in accordance with Technical p Specification 3.3.3.6. With less than the minimum number of L channels operable and reasonable efforts to return the instrument (s) to operable status within 30 days being unsuccessful, Technical Specification 3.3.6.b requires an explanation for the delay in correcting the inoperability in the

[ next Semiant.ual Ef fluent Release Report.

Response: Since the requirements of Technie.1 Specification 3.3.3.6

{ governing the operability of radioactive liquid effluent mon'.toring instrumentation were met for this reporting period, c' response is required.

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L APPENDIX B Radioactive Caseous Effluent Monitoring Instrumentation ,

Requirement: Radioactive gaseous effluent monitoring instrumentation e.hanncis are required to be operable in accordance with Technical Specification 3.3.3.7. With less than the minimum number of channels operable and reasonable efforts to return the I instrument (s) to operable status within 30 days being unsuccessful. Technical Specification 3.3." 7.b requires an explanation for the delay in correcting the inoperability in the

. ext Semiannual Ef fluent Release Report.

Response: Since the requirements of Technical Specification 3.3.3.7 governing the operability of radioactive gaseous effluent monitsring instrumentation were met for this reporting period, no response is required.

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B-1 4824R/20.124

I L APPENDIX C Liquid Holdup Tanks Requirement: Technical Specification 3.11.1.4 limits the quantity of radioactive material contained in any outside temporary tank.

With the quantity of radioactive material in any outside temporary tank exceeding tr.e limits of Technical Specification I 3.11.1.4, a description of the events leading to this condition is required in the next Semiannual Effluent Release Report.

I Response: The limits of Technical Specification 3.11.1.4 were not exceeded during this reporting period.

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C-1 2824R/20.124

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[ APPENDIX D Radiological Environmental Monitoring Program p Requirement The radiological environmental monitoring program is conducted

• in accordance with Technical Specification 3.4.12.1. With milk or freuh leafy vegetation samples no longer available from one or more of the required sample locations. Technical Specification 3.12.1.c requires the identification of the new location (s) for obtaining replacement sample (s) in the next Semiannual Effluent Release Report and inclusion of revised Off-Site Dose Calculation Manual Figure (s) and Table (s)

{ reflecting the new location (s).

Response: No new sampling locations were needed to be identified in accordance with the above requirement.

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APPENDIX E

( I.and Use Census Requirement: A land use census is conducted in accordance with Technical

( Specification 3.12.2. With a land use census identifying a location (s) which yields at least a 20 percent greater dose or dose comitment than the values currently being calculated in Technical Specification 4.11.2.3. Technical Specification 3.12.2.a requires the identification of the new location (s) in the next Semiannual Effluent Release Report.

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Response: The annual Land Use Census was not completed during this reporting period.

(

Requirement: With a land use census identifying a location (s) which yields a -

calculated dose or dose comitnxat (via the same exposure

- pathway) at least 20 percent greater than '. a location from L which samples are currently being obtaire) in accordance with y

Technical Specification 3.12.1. Teche.ical Specification 3.12.2.b L requires the identification of the new location (s) in the next Semiannual Effluent Release Report.

e Respor.se' The annual Land Use Census was not completed during this reporting period.

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APPENDIX F Process Control Program Requirement: Technical Specification 6.14.1 requires that licensee initiated changes to the Process Control Program be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made.

Response: There were no licensee initiated changes to the Prn;ess Control Program during this reporting period.

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1 F-1 4824R/20.124 l

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L AFPENDIX C Off-Site Dose Calculation Manual

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Requirement: Technical Specification 6.15.2 requires that licensee initiated changep to the Off-Site Dose Calculation Manual be submitted to

'I the Commission in the Semiannuol Radioactive Effluent Release Report for the period in which the change (s) was made effective.

Responce: There was one licensee initiated revision to the Off-Site Dose Calculation Manual during this reporting period which consisted of the following:

1 1. Section 4.0 of the ODCM was revised in Revision 6 to reflect the following changes:

a. Food product sampling locstion IF-12 was deleted from I Table 4.1 since samples from there were no longer required after October 31, 1986.

b. The fenceline location and several building names and locations were updated in Figure 4-4.

2. The ODCM was also revised in Revision 6 to reflect I additional changes in the liquid dose and gaseous dose and dose rate calculations. These changes are identified as I follows:

a. The liquid dose factors in Section 3.0 were revised to reflect the additional dose pathways of irrigated food crops to man, and irrigated food crops to cow to man.

These additional pathways were included as r. result of a survey which indicated that water from the Deerfield i River was used to irrigate food crops.

G-1 4824R/20.124 i

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i l APPENDIX G (Continued) l

b. The liquid dose equations in Section 3.0 were revised to incorpcrate a Deerfield River flow rate correction f actor to account for variation in the river flow rate I during the year. The minimum monthly ten-year average (August 1974 to September 1984) Deerfield River flow I rate below Sherman Dam was used in the correction factor as a default value if actual river flow date during release periods is unavailable.

c. In the example calculations in Section 5.0, the steam generator flow rate, f , 2was changed to 100 gpm to reflect the maximum flow rate durir.g actual discharge, and generate the most restrictive liquid monitor setpoint.

I

d. The text in Section 6.0 was revised to reflect a flow rate meter placed on the steam generator blowdown tank discharge line to estimate the discharge rate during periods of release. The flow rate meter was installed during the previous reporting period in response to Technical Specification Amendment Change 92 which required a flow rate measuring device for the steam generator blowdown tank effluent (Table 3.3-8).

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e. The noble gas dose equations, dose conversion factors for individual iodine, tritium, and particulate radionuclides, description of bases for equations in Section 3.0, and effluent monitor setpoint equations in Section 5.0, were revised for a more recent 5-year (1981 through 1935) period than the original ODCM.

I G-2 1 AS24R/20.124 l

l APPENDIX G (Continued) l

f. The iodine, tritium, and particulate dose and dose rate factors in Section 3.0 were expanded to include nuclides not considered previously in the dose and dose rate I calculations,

g. The noble gas dose rate equations and dose rate factors for iodine and particulate radionuclides in Section 3.0 were revised to change the shielding factor from 0.7 to 1.0 to account for the fact that the dose rate applies at all times at, or beyond, the site boundary.

h. Section 3.0 of the ODCM was also expanded to include a methodology for calculating the gamma air and beta air doses in 'he case of an unexpected ground level noble i gas release, since other listed gaseous doce equations relate to the plant stack only.

i. The following changes relate to the calculational development of the dose factors in Section 3.0:

(1) The agricultural productivity for pasture grass was 1 changed from 0.75 kg m - to 0.7 kg m- to keep consistent with USNRC Regulatory Guide 1.109 I Revision 1, October 1977.

(2l The crop exposure time to the plume for stored feed was changed from 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> to 1440 hours0.0167 days <br />0.4 hours <br />0.00238 weeks <br />5.4792e-4 months <br />, since the stored feed that cattle consume at and around Rowe is corn silage.

I (3) The fraction of leafy vegetables grown in the garden of interest was changed from 0.5 to 1.0 to I keep consistent with Regulatory Guide 1.109.

G-3 4824R/20.124

APPENDIX G (Continued)

(4) The fraction of iodine assumed to be in elemental form was corrected from 1.0 to 0.5 to keep consistent with Regulatory Guide 1.109.

(5) The absolute humidity value was changed from

~

to 5.6 g m

-3 since it is a more 8.0 g m representative average value for the region. The value of 5.6 g m~ was taken from Health Physics -

Journal, Volume 39, Number 2, August 1980, Pergammon Press, Ltd.

(6) The ground deposition factors (D/Q values) which are factored into the critical organ dose and dose rate factors were revised to reflect the relative I deposition rate curves from USh7C Regulatory Guide 1.111. Revision 1 July 1977.

J. Misce!'sneous typographical errors were also corrected throughout the ODCM which had no impact on the methodology or parameter values used.

I None of the above changes 'dll reduce the accuracy or reliability of dose calculations or setpoint determinations.

I The calculated methodology is the same as in the original version of the ODCM and remains consistent with Regulatory Guide 1.109 changes. Changes in the dose eticulations relate to parameter values which have been updated to reflect Regulatory Guide 1.109 values or more conservative site-related values.

The above changes have been reviewed and approved by PORC.

The revised ODCM pages reflecting the above changes are included I as part of this Appendix.

G-4 4824R/20.124

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YMa'EE NUCLEAR PO'a'ER STATION n OFF-SITE DOSE CALCULATION MANUAL l YANY,EE ATOMIC ELECTRIC COMPANY NUCLEAR SERVICES DIVISION 1671 WORCESTER ROAD ThMtISGHM MASSACH'JSETTS 01701 1

I I

PitEPARED BY/DATE REVIDED BY/MT11 APPPO\T.D Andrew D. Hodgdon 12/2/82 ORIGINAL William D. Billings 12/2/82 PORC 11/29/82 RFYlSlON } [&&f ^ V /f/0 k fNl?f

-- _ ~/r 3f3'fW Marh.5o s isty 85 -3 REVISION 2 EM M LE r/7fs $U{. "*n1"h g f, # ' l' 7 , 19 REVISION 3 g/g PORC Meeting 86-16 Y #b

, /f ,. . - 'y' 'a -' ' '- / ' '

February 25, 1986 ///.' f/ [ui:/

PORC Meetin6 06-' -

W 't-REvlslON 4 f5,,c<<</ 4 64 , .,, . 7 r/a M -

May 20, 198 c. ,

a...m.,

s .

NRC Meetin6 06-59 '

RtvlS10N 5 f, . . n / N ' d. , . . . . . 7 5 '2 ' f" Sept. 30,1986jj, ,;, c, ,'[, . -

/ FORC I..eeting ~ eo-Oc -

REVISIOS 6 , , ,, , ,

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L REVISION RECORD 1

1 Revision Date Description 1 0 12/1/82 Initial printing. Approved by PORC 11/29/82.

Submitted for USNRC approval 12/3/82, 1 3/30/84 Change in environmental monitoring sampling locations based on 1983 land use census. Errors in Table 4.1 corrected. Haps revised.

2 7/30/85 Addition of Intercomparisor. Program description to Section 4.0. Reviewed by PORC 7/30/85.

3 3/19/86 Addition of a PVS I-131 inspection limit to demonstrate compliance with Technical Specification 3.11.2.1.b.

4 5/21/86 Change in milk satcling location. Samples no longer available at Station TM-II.

5 9/30/86 Change in food product sampling location based on 1986 land use census.

6 2/18/88 Change in liquid dose factors to reflect additional dose pathways. Change in gaseous dose factors to reflect five-year average meteorology. Change in I gaseous dose rate factors to reflect a shielding factor of 1.0. Deletion of food product location TF-12 (samples no longer required after 10/31/86).

i Update of fence line location and seve.'al building names and locatio'ts in Figure 4-4.

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REVIS!ON RECORD

[ LIST OF AFFECTED PAGES Changes, deletions or additions in the most recent revision are indicated by a bar in the margin or by a dot near the page number if the entire page is affected.

l Page Revision i to 111 6 tv to vii, ix 6 vili 1 4-1 to 4-10 1 4-1 2 3-24 to 3-26 3 4-3, 4-9 4 4-3, 4-9, 4-10 5 1-1 to 1-17 6 3-1 to 3-50 6 1

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DISCLAIMER OF RESPONSIBILITY E

This document was prepared by Yankee Atomic Electric Company ("Yankee"). The use of information contained in

( this document by anyone other than Yankee, r)r the Organization for which the document was prepared under contract, is not authorized and, with respect to any unauthorized use, neither Yankee nor its officers,

{ directors, agents, or employees assume any obilgation, responsibility, or liability or make any warranty or representation as to the accuracy or completeness of the

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ABSTRACT The YNPS OOCH (Yankee Nuclear Power Station Off-Site Oose Calculation r Manual) contains the approved methods to estimate the doses and radionuclide concentrations occurring beyond the boundarles of the plant caused by normal plant operation. With initial approval by the U.S. Nuclear Regulatory Commission and the YNPS Plant Operation Review Committee (PORC) and approval of subsequent revisions by PORC (as per the Technical Specifications) this OOCH is suitable to show compliance where referred to by the Plant Technical

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Specifications, b

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L TABLE OF CONTENTS b em ii REv!SiON REC 0RD..................................................

g 111 LIST OF EFFECTIVE PAGES..........................................

iv

( DISCLAIMER OF RESPONSIBILITY.....................................

v ABSTRACT.........................................................

[ u S T O r n GU R t S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vm i,

uS1OrusuS...................................................

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i-i i.0 m R0 DUCTION.....................................................

( 1.1 Summary of Methods. Dose Factors. Limits. Constants.

Variables and Definitions.................................. 1-2 2.0 METHOD TO CALCULATE OFF-SITE LIQUID CONCENTRATIONS............... 2-1

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2.1 Method..................................................... 2-1 2.2 Method to Determine Radlonuclide Concentration

[ for Each Liquid Effluent Pathway........................... 2-2 2.2.1 Test Tank Pathway.................................. 2-2 2-3

[ 2.2.2 2.2.3 Steam Generator Blowdown Pathway...................

Secondary Coolant and Coolant Leakage Pathway...... 2-3 2.2.4 Remaining Pathways................................. 2-3 2.3 Background Information..................................... 2-4 3.0 0FF-SITE DOSE CALCULATION METH005................................ 31 3.1 Introductory Concepts...................................... 3-2 3.2 Method to Calculate Total Body Dose From Liquid 3-5

[ 3.3 Releases...................................................

Method to Calculate Maximum Organ Dose From Liquid I

Releases................................................... 3-11 3.4 Method to Calculate the Total Body Dose Rate From

[ 3.5 Noble Gases................................................

Method to Calculate the Skin Dose Rate From Noble Gases....

3-14 3-19 1

3.6 Method to Calculate the Critical Organ Dose Rate From

( 131-1, 3H and Particulates with T1/2 Greater Than 8 Days..................................................... 3-23 3.7 Method to Calculate the Gamma Air Dose From Noble Gases.... 3-27 3.8 Method to Calculate the Beta Air Dose From Noble Gases..... 3-31

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TABLES OF CONTENTS (Continued) 3.9 Method to Calculate the Critical Organ Dose from Tritium, Iodines and Particulates................................... 3-34 3.10 Critical Receptors and Annual Average Atmospheric Dilution Factors for Important Exposure Pathways........... 3-41 3.11 Method to Calculate Direct Dose From Plant Operation....... 3-45 4.0 ENVIRONMENTAL MONITORING LOCATIONS............................... 4-1 5.0 SETPOINT DETERMINATIONS.......................................... 5-1

[ 5.1 Liquid Effluent Instrumentation Setpoints..................

Gaseous Effluent Instrumentation Setpo1nts.................

5-2 5-7 5.2 b 6.0 LIQUID AND GASEOUS EFFLUENT STREAMS RADIATION MONITORS AND RADWASTE TREATMENT SYSTEMS................................... 6-1

[ 6.1 6.2 In-Plant Liquid Effluent Pathways..........................

In-Plant Gaseous Effluent Pathways.........................

6-1 6-3 R-i

[ RErERENcES.......................................................

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11 LIST OF TABLES 1

Number Title Page 1.1-1 Summary of Concentration and Setpoint Methods, and Method I Dose Equations for Normal Operations at Yankee Plant 1-3 1.1-2 Dose Factors Specific for Yankee Plant for Noble Gas Releases 1-7 1.1-3 Summary of Radioactive Effluent Technical Specifications With Dose or Dose Rate Limits and Implementating Method I Equations 1-8 1.1-4 Summary of Constants 1-10 l 1.1-5 Summary of Variables 1-11 1.1-6 Definition of Terms 1-15 1.1-7 Dose Factors Specific for Yankee Plant for Liquid Releases 1-16 1.1-8 Dose and Dose Rate factors Specific for Yankee Plant i for Tritium, Iodine, and Particulate Releases 1-17 I 2-1 Typical Radionuclides Released From Test Tanks 2-6 3.2-1 Environmental Parameters for Liquid Effluents at Yankee Plant 3-9 3.2-2 Age Specific Usage Factors for Various Liquid Pathways 1 at Rowe 3-10 3.9-1 Age Specific Usage Factors 3-38 3.9-2 Environmental Parameters for Gaseous Effluents at Yankee Plant 3-39 F 3.10-1 Yankee Nuclear Power Station Five-Year Average Atmospheric Dispersion factors 3-44 m 3.11-1 Estimate of Exposure Rate at Critical Receptor From VC Shine, EVC, Made in Spring 1981 3-50 4-1 Radiological Environmental Monitoring Stations 4-13 5.2-1 Sample Caleviations of Gaseous Effluent Satpoint 5-12 5.2-2 Relative Fractions of Core Inventory Noble Gases After Shutdown 5-13

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1.0 INTROOL; TION

[ The OOCH (Off-Site Dose Calculation Manual) provides formal and approved methods for the calculation of off-site concentration, off-site doses and effluent monitor setpoints, and indicates the locations of environmental monitoring stations in order to comply with the Yankee Nuclear Power Station b Radiological Effluent Technical Specifications (RETS) Sections 3/4.3.3.6, 3/4.3.3.7, and 3/4.11, as well as the Radiological Environmental Monitoring

[ Program (Section 3/4.12). The ODCH forms the basis for plant procedures which document the off-site doses due to plant operation which are used to show compliance with the numerical guides for design objectives of Section II of

{ Appendix I to 10CFR Part 50.

The methods contained herein follow accepted NRC guidance, unless otherwise noted in the text. The basis for each method is sufficiently documented to allow regeneration of the methods by an experienced Health Physicist.

All changes to the ODCM shall be reviewed and approved by the Plant Operation Review Committee (PORC) in accordance with Techrlical Specification

{ 6.15 prior to implementation. Changes made to the ODCM shall be submitted to the Commission for their information in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective.

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F-L Summary of Methods. Dose Factors. Limits. Constants. Variables and 1.1 r Definitions L

This section summarizes the methods for the user. The first time user should read Chapters 2 through 5. The concentration and setpoint methods are i documented in Table 1.1-1, as well as the Method I Dose equations. Where more

[ accurate Dose calculations are needed use the Method !! for the appropriate dose as described in Sections 3.2 through 3.9 and 3.11. The dose factors used in the equations are in Table 1.1-2, 1.1-7, and 1.1-8 and the Regulatory .l

{ Limits are summarized in Table 1.1-3.

b The constants, variables, and special definitions used in this ODCM are in Tables 1.1-4, 1.1-5, and 1.1-6.

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Table 1.1-1 Summary of Concentration and Setpoint Methods, and Method I Dose Equations for Normal Operations at Yankee Plant

[ Equation No, Maximum Equatton j

ENG 2-1 Unrestricted Area FfNG,3 Total Fraction of MPC in u i Liquids Ixcept Noble Gases

{

NG =

2-2 Unrestricted Area Concentration C Cf of Noble Gases in Liquids 3-1 Total Body Dose from Liquids Dtb(**} " K f0DFL 1 itb I b 3-2 Organ Dose frcm Liquids D organ (*"*)*5f0DFL gg 1 3-3 Total Body Dose DFB g Rate from Noble Gases btbI'Yr } " I'83 1 3-4 Skin Dose Rate from Noble Gases bskin( '}*I r i 1 DFj j

3-5 Organ Dose Rate from 1311, H3 and g gerem) =

Particulates with Tl/2 > 8 days to yr fih DFG'co t E

3-6 Gamma Air Dose from D tr(mrad)=0.25{Qg DF{

Noble Gases

{

3-7 Beta Air Dose from Dfir(mrad)=0.76IQDFf g Noble Gases i  !

3-6.1 Gamma Air Dose from Dgrd(mrad) = (1,23E-04)(Q y ,,j33 l Ground Level Noble Gas equivalent)

Releases

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Table 1.1-1 (continued)

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Summary of Concentration and Setpoint Methods, and Hethod I Dose Equations for Normal Operations at Yankee Plant Equation g No. Maximum Equation

( i 3-8 Organ Dose from 1311, 3H and Q DFG ico Particulates with T1/2 > 8 days Den (mrem) - ,

(

3-9 Direct Dose Dd - (0.057 + d r) T, 0.00087

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5-1 Liquid Release Rate Reading Rf ,f HPC g Sg

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S( ffs)(500)60

( Gaseous Release Rate Reading R g g 5-3 tb "

for Total Body Dose Limit F7.83{fh^DSF g L

NG sg)(3000) 60 Gaseous Release Rate Reading R S ({ f g

5-4 sk "

for Skin Dose Limit F If"GDFj i

Note 1:

Cg - Concentration radionucilde 1 in mixture (pct /ml).

p L ,

E - Exposure rate at critical receptor from non-VC sources as y r measured or estimated for the period (pR/hr).

L F = Primary Vent Stack Flow Rate (cc/ min).

MG = Fraction activity of radionuclide i to total noble gLs

[ f activity.

"1", except noble gases, at point F

L Cf"O -ofConcentration discharge. of radionuclide C"O - Concentration of radionuclide "t", except noble gases, at point

( of discharge.

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E L. Table ).1-1 (continued)

Summary of Concentration and Setpoint Methods, and Method I Dose Equations for Normal Operations at Yankee Plant OF' . Skin dose factor for radionuclide "i".

OFJ = Gamma dose factor to air for radionuclide "i".

OF - Beta dose factor to air for radior.uclide "i".

F OFBg . Total body dose factor for radionuclide "1".

OFGggo Site-specific, c itical organ dose factor for a gaseous release of nucl'de "i".

OFGlg-Site-spe:ificcriticalorgandoseratefactorfora "1".

gaseous release cf nuclide S te-specific total Lody dose factor for a liquid OFLitb release of nuclide "1". {

l DFLI *# = Site-specific, tranimum organ dose f actor for a liquid release cf nuclide "1". l Flow rate past test tank nonitor (gpm).

f) f.y . Flow rate past steam generator blowdown monitor (gpm),

f Flow rate at point of discharge (gpm).

3 I

K . Deerfield River flow rate correction factor.

HPC , Composite HPC for the mix of radionuclides (pCi/ml),

2C g

- (Eq. 5-2)

I _Ci_

i MPC g Qg . Total release (curies) for radionuclide "1".

9 Xe-133 equivalent Total release of noble gases expressed as Xe-133 equivalent.

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Iable 1.1-1 -

(continued)

" Summary of Concentration and Setpoint Methods, ana Method I Dose Equations for Normal Operations at Yankee Plant a s hg - Release rate (UC1/sec) for radionuclide "i".

S - Gaseous Instrument response factor (cpm /(pC1/cc)).

g Sg - Ligt'il instrument response factor (cpm /(pC1/cc)).

sg - Ratio e' ~ % from equal activities of radionuclide i to a refert  %..aciido, i.e., Xe-133. ,

T, = Length of exposure period (hours).

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!able 1.1-2 Dose Factors Specifit for Yankee Plant for Noble Gas Releases a

k Gamma

'g Total Body Beta Skin Combined Skin Beta Air Gamma U-Dose Facto Dose Factor D se Fact; Dose Factor DoseFacto5

*-*3 DFS(*#'*-*g DF'i (mrem-sec) 0F (mrad-m ) DFY (r6v 1

r<adionucl i de DFS i (*pCi-yr ) i DCi-yr ) pCi-yr DCi-yr i pCi-y-2.69E-03 1.45E-01 3.28E-03 9.30E-0; Ar-41 8.84E-03*

Kr-83m 7.56E-08 ----- 1.68E-04 2.88E-04 1.93E "

Kr-85m 1.17E-03 1.46E-03 4.56E-02 1.97E-03 1.23L-::

Kr-85 1.61E-05 1.34E-03 3.22E-02 1.95E-03 1.72E-::

J Kr-87 5.92E-03 9.73E-03 2.86E-01 1.03E-02 6.17E-::

e Kr-88 1.47E-02 2.37E-03 1.89E-01 2.93E-03 1.52E-C' 1.66E-02 1.01E-02 3.92E-01 1.06E-02 1.73E-C; Kr-S9 1.56E-02 7.29E-03 3.16E-01 7.83E-03 1.63E-0; Kr-90 Xe-131m 9.15E-05 4 76E-Oi 1.27E-02 1.11E-03 1.56E Xe-133m 2.51E-04 9.94E-04 2.66E-02 1.48E-03 3.27E Xe-133 2.94E-04 3.06E-04 1.04E-02 1.05E-03 3.53E-C Xe-135m 3.12E-03 7.11E-04 4.62E-02 7.39E-04 3.36E-0;-

Xe-135 1.81E-03 ' 86E-03 6.11E-02 2.46E-03 1.92E-C'

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Xe-137 1.42E-03 1.22E-02 3.05E-01 1.27E-02 1.SiE-0.5 Xe-138 8.83E-03 4.13E-03 1.79E-01 4.75E-03 9.21E-02 8.84E 8.84 x 10-3 I

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" Table 1.1-3 7 Summary of Radiological Effluent Technical Specificationt L, and Implementing Equations Technical Specification Category Method

• _,

Limit 3.3.3.6 Liquid Effluent Alarm / Trip Setpoint Eq. 5-1 T.S. 3.11.1.1 Honitor Setpoint 3.3.3.7 Gaseous Effluent Alarm Setpoint for Eq. 5-3 T.S. 3.11.2.la Honitor Setpoint Total Body Dose Rate (Total Body)

Alarm Setpoint for Eq. 5-4 T.S. 3.11.2.la Skin Dose Rate (Skin) 3.11.1.1 Concentration Total fraction of Eq. 2-1 1 1.0 '

(Liquids) HPC Excluding Noble Gases I Total Noble Gas Eq. 2-2 1 2x10-4 pCi/cc Concentration 3.11.1.2 Dose (Liquids) Total Body Dose Eq. 3-1 1 1.5 mrem in a qtr.

1 3.0 terem in a yr.

l Organ Dose Eq. 3-2 < 5 mrem in a qtr.

I 10 mrem in a yr.

l 3.11.1.3 Liquid Radwaste Total Body Dose Eq. 3-1 1 0.06 mrem in a mo.

Treatment Organ Dose Eq. 3-2 1 0.2 mrem in a mo.

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3.11.2.1 Gaseous Effluents Total Body Dose Rate Eq. 3-3 1 500 mrem /yr.

Dose Rate from Noble Gases Skin Dose Rate from Eq. 3-4 1 3000 rrem/yr.

Noble Gases j Or,in Dose Rate from Eq. 3-5 1 1500 mrem /yr.

1131, H3, < id Particulates with Tl/2 >8 Days l

/ s' Revision 6 - 2/18/88 Approved By:4f:I/C-

L Table 1.1-3

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L Summary of Radiological Effluent Technical Specifications and Implementing Equations Technical Specification Category Method

• Limit Eq. 3-6 1 5 mrad in a qtr.

i 3.11.2.2 Gaseous Effluent Gamma Air Dose from Dose, Noble Gases Noble Gases 1 10 mrad in a yr.

Beta Air Dose from Eq. 3-7 1 10 mrad in a otr.

Noble Gases 1 20 mrad in a yr. ,

3.11.2.3 Gaseous Effluent Organ Dose from Eq. 3-8 1 7.5 mrem ir, a qtr.

Dose, lodine-131, 1131 H3, and Tritium and ParticJiates with 1 15 mrem in a yr.

I Radionuclides Tl/2 >8 Days 3.11.2.4 Gaseous Redweste Gam:Ta Air Oose from Eq. 3-6 1 0.2 mrad in a me. j I Treatment Noble Gases l Beta Air Oose from Eq. 3-7 1 0.4 mrad in a mo.

N0ble Gases 1

Organ Dose from 1131, Eq. 3-8 10.3 mrem in a mo.

H3, and Particulatos I with T]/2 >6 days 3.11.4 Total Dose from Total Body Ocse Eq. 3-l+ 1 25 mrem in a yr.

All Sources Organ Dose Eq. 3-6+

1 Thyroid Dose Eq. 3-9 Eq. 3-2+ 1 25 mrem in a yr.

Eq. 3-8+

Eq. 3-9 ,

Eq. 3-2+ 1 75 mr3m in a yr.

Eq. 3-8+

Eq. 3-9 1

• More accurate methods may be available (see s'ibsequent chapters).

I p Revision 6 - 2/18/8s Approvee Sy s g M L l-9 5003R/26.219

Table 1.1-4 Summary of Constants

, Constant- Definition e 's

( 0.00087 - Conversion factor mrem pR

( 0.25 3.17 x 10 4(h 3

h) (X/Q)Y (sec/m )

4 DC1-( = (3.17 x 10 )(7.83 x 10-6)

Ci-m 3

- 0.76 3.17 x 104([ h) (X/Q) (sec/m )

DCl-F r - 3.17 x 104 (2.39 x 10-5) 3 L C1-m 1.11 Average ratto of tissue to air energy

( absorption ratto coefficient 7.83 x 10-6 (3,ef,3) pCi-sec H 7.83 - 106 (pCi/uC1)

• 1.0 3 uCi-m l L

8.69 - 1.11 57 (X/0]Y (sec/m ) 1 x 100 (pCl/ C1)

DCl-sec

- 1.(1.0)(7.83 x 10-6)(1 x 106 ) 3 3 C1-m 23.9 - 1 x 106 (X/Q)

= (1 x 106)(2.39 x 10-5) r 60. Conversion factor sec min

[

500. . Total body annual dose limit from ICRP2 mrem 3000. Skin annual dose limit from ICRP2 mrem DC1-sec I

3.I'/ x 10 4

- Number of picoeuries per curie divided by Cl-yr number of seconds per year I

( ) x 10 0 Number of picocuries per microcurie Revision 6 - 2/18/88 Approved By:M

{

~ 1-10 L 5003R/26.219

Table 1.1-5 Summary of Vartibles

(

Variable , Definition Units pCl C"O = Total activity of all dissolved and entrained CC noble gases from all station sources C

ENG - Concentration 6f radionucilde "1", except cc I' noble gases, at point of discharge (l

uCi C

NG

- Concentration of radionuclide "l", except CC I noble gases, at point of discharge

(

3 C - Concentration of radionuclide "1" uC1/m I or pC1/cc

(

6 mrad D o Beta dose to air 3r D - Gamma dose to air mrad alr L

O - Gamma dose to air from a ground level release mrad l grd D - Dose to the critical organ mrem en

[ D - Direct dose mrem d

D - Dose to the maximum organ mrem organ D a Do',e to skin from beta and gamma mrem skin l-D - Cose to the total body mrtm tb J

l 3

' Total body gamma dose factor for nucilde "1" DFB g -

{

" 3 DFS g - Beta skin dose factor for nuclide "1" {

= Combined site-specific skin dose factor u r DF{

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L Table 1.1-5 (continued)

Summary of Variables Variable Definition Units ad

- Gamma air dose factor for nuclide "1" DFJ _

0 F DF - Beta air dose factor for nuclide "1"

, DFG gC0

- Critical organ gaseous dose factor for mrem nuclide "1" Ci i DFG'CO I

- Critical organ gaseous dose rate factor fcr nuclide "i" mrem-sec pCi-yr

= Naximum organ liquid dose factor for mrem I DFLI *0 nuclide "1" Ci DFl - Total body liquid dose tactor for mrem i itb I

nuclide "i" Ci i

b - Critical organ dose rate due tt Iodines, yr I CO tritium, and particulates b - Skin dose rate due to noble gases

• f*

skin b

tb

- Total body dose rate due to noble gases *f*

D/Q - Deposition factor for dry deposition of ge I elemental radiolodines and other particulates m 2

E - Exposure rate at critical receptor from pR r

I non-VC sources as measured or estimated for the period hr E

VC

- Lim ting exposure rate at the critical pR 8 receptor from the Vapor Container during hr rormal operations i F - Primary vent stack flow rate h C

f - Fractl6n activity of radionuclide i to total noble gas activity /

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5003R/26.219

Table 1.1-5 (continued)

St.mmary of Variables Variable Definition Units

- Total fraction of MPC in liquid pathways F)

F ENG - Total fraction of MPC in liquid pathways I (excluding noble gases) f j

- Flow rate past test tank monitor gpm f - Flow rate past steam generator monitor gpm 2

f - Flow rate at point of discharge gpm 3

- Composite MPC for the mix of radionuclides "'

MPC cc C

(see Equation 5-2)

HPC Maximum permissib'e concentration radionuclide $

I cc "i" (10CFR20, Appendix B, Table 2, Column 2)

Q - Total release of all noble gases Curies Qg - Release for rudionuclide i Curies I - Total release rate of all noble gases UCuries/sec h

hg - Release rate for radionuclide "1" uCuries/sec X/Q - Average undepleted dispersion factor m

(X/Q)0 - Average depleted dispersion factor m

S (X/Q)T - Effective average gamma dispersion factor m$C 5 - Shielding factor Ratio 7

5 - Gaseous monitor response factor cc 9

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I' u

Table 1.1-5 m Icontinued)

L Summary of Variables b Definition Units Variable sg - Ratio of response from equal activities of

{' radionuclide i to a reference radionuclide (such as Xe-133)

Sg - Liquid monitor response factor Ci

[ T, AD

- Exposure period

- Conservative increment in annual average dose hours mrem

.9

[

[

[

[

[

[

[

[

[

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( _ - - - - - - - - - - .

Table 1.1-6

[ Oefinition of Terms I

I Table De'.eted.

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TABLE 1.1-7 Oose Factors Specific for Yankee Plant

[. for

~

Liquidiieleases

[ Total Body Maximum Organ Dose Factor Dose Factor

* I'*

(- Radlonuclide 0FL itb (*Ci )

DFL imo ("C1 )

I H-3 5.99E-04 5.99E-04 -

C-14 1.64E+00 8.18E+00 Cr-51 7.20E-05 1.07E-02 Mn-54 6.07E-02 5.47E-01

{ Fe-55 3.46E-02 2.11 E-01 Fe-59 1.00E-01 4.53E-01 Co-58 4.76E-02 1.81E-01 p 9.04E-01 L Co-60 2.79E-01 Zn-65 1.65E+00 2.71E+00 Sr-89 2.30E-01 8.04E+00 I Sr-90 6.97E+01 2.75E+02

' 1.40E-03 2.87E-01 Zr-95/Nb-95 Ru-103 2.48E-03 3.57E-01 p Ag-110m 2.32E-02 2.21E+00 L Sb-124 2.62E-02 6.48E-01 I-131 8.57E-03 4.96E+00 1-133 6.52E-04 3.18E-01

[ Ci-154-Cs-136 1.79E+01 2.28E+00 2.40E+01 3.20E+00 Cs-137 1.076+01 2.07E+01 I Ba-140/La-140 3.40E-03 5.80E-02 L Ce-141 7.73E-05 1.06E-01 Ce-144 1.41E-03 2.58E+00 E

E E

E

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TABLE 1.1-8 Dose and Dose Rate Factors Specific for Yankee Plant LoE 3a lodines, Tritium and Particulate Releases Critical Organ Critical Organ Dose Factor Dose Rate Factor Radionuclide DFG ico (*Ci ) DFG'co i (mrem-sec) yr-pCi H-3 7.21E-03 2.27E-01 C-14 4.38E+00 1.38E+02 1 Cr-51 3.44E-02 1.19E+00 Hn-54 3.78E+00 1.49E+02 Fe-59 3.83E+00 1.27E+02 I Co-58 Co-60 1.98E+00 4.08E+01 7.06E+01 1.812+03 Zn-65 1.99E+01 6.43E+02 I Sr-89 Sr-90 6.10E+01 2.36E+03 1.92E+03 7.44E+04 Zr-95/Nb-95 3.77E+00 1.24E+02 i Ru-103 Ag-110m Sb-124 1.02E+01 3.63E+01 6.95E+00 3.22E+02 1.22E+03 2.32E+02 i 1-131 4.19E+02 1.32E+04 i I-133 Cs-134 6.29E+00 8.52E+01 1.98E+02 2.83E+03 Cs-136 4.71E+00 1.52E+02 8.71E+01 2.97E+03 l Cs-137 Ba-140/La-140 1.44E+00 4.60E+01 Ce-141 9.75E-01 3.10E+01 Ce-144 2.10E+01 6.65E+02 1

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I 3.0 0FF-SITE DOSE CALCULATION METHODS F

Chapter 3 provides the basis for plant procedures that the plant

- operator requires to meet the Dose Radiological Effluent Technical Specifications (hereafter called Dose RETS). A simple, conservative method (called Method I) is listed in Table 1.1-1 for each of the nine doses required by the Dose RETS. Each of the nine Method I eauations is presented, along with their bases in Sections 3.2 through 3.9 and Section 3.11. In addition, those sections include more sophisticated methods (called Method II) for use when more accurate results are needed. This chapter provides the methods, l data, and reference material with which the operator can calculate the needed doses.

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3.1 Introductory Concepts The Radiological Effluent Technical Specifications (RETS) either limit dose or dose rate. The term "Dose" for ingested or inhaled radioactivity means the dose commitment, measured in mrem, which results from the exposure to radioactive materials that, because of uptake and deposition in the body, will continue to expose the body to radiation for some period of time after the source of radioactivity is stopped. The time frame over which the dose commitment is evaluated is 50~ years. The phrases "annual Dose" or "Dose in one year" then refers to the fifty-year dose commitment from one year's worth of releases. "Dose in a quarter" similarly means a fifty-year dose commitment from one quarter's releases. The term "Dose," with respect to external I exposures, such as to noble gas clouds, refen only to the doses received l during the actual time period of exposure to the radioactivity released from the plant. Once the source of the radioactivity is removed, there is no longer any additional accumulation to the dose commitment.

I Gaseous effluents from the plant are also controlled such that the maximum "dose rates" at the site boundary at any time are limited to 500 mrem /yr to the whole body or 3000 mrem /yr to the skin. The annual dose limits are the doses associated with the concentrations of Appendix B, Table I 11, Column 1 of 10CFR Part 20 (10CFR20.106(a)). The use of the annual dose l limits embodied in 10CFR Part 20 as plant "dose rate" values (to be applied at any time consistent with the capabilities of the monitoring instrumentation to determine) provides reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of member (s) of the public either within or outside the site boundary to annual average concentrations exceeding the federal regulations.

It should also be noted that a dose rate due to noble gases that exceeds, for a short time period (less than one hour in duration), the equivalent 500 mrem / year dose rate limit stated in Technical Specification 3.11.2.1, does not necessarily by itself constitute a Licensee Event Report (LER) under 10CFR Part 50.73 unless it is determined that the air concentration of radioactive effluents in unrestricted areas has also exceeded p f /  !

Revision 6 - 2/18/88 Approved By: 4 7'd f / di'A 5008R/12.332 3-2

two times MPC when averaged over one hour (four-hour notification per

- 10CFR50.72, and 30-day LER per 10CFR50.73).

1 ThequantitiesDandbareintroducedtoprovidecalculablequantities, related to off-site dose or dose rate which demonstrates compliance with the RETS.

The dose D is the quantity calculated by the Chapter 3 dose equations.

The O calculated by "Method I" equations is not necessarily the actual dose received by a real individual but usually provides an upper bound for a given release because of the conservative margin built into the dose factors and the selection and definition of critical receptors. The radioisotope specific dose factors in each "Method I" dose equation represent the greatest dose to any organ of any age group accounting for existing or potential pathways of exposure. The critical receptor assumed by "Method I" equations is typically a hypothetical individual whose behavior - in terms of location and intake -

results in a dose which is expected to be higher than any real individual, j Method II allows for a more exact dose calculation for real individuals, if necessary, by considering only existing pathways of exposure, or actual concurrent meteorology with the recorded release.

bisthequantitycalculatedintheChapter3doserateequations. It is calculated using the plant's effluent monitoring system reading and an annual average or 'ong-term atmospheric dispersion factor. If plant release rates were such that a 0 equal to the Technical Specification (3.11.2.1) value I was continued for one year, the annual dose limits of 10CFR20 would be reached. However, since maximum allowed release rates and the resulting dose rates in the range of the Technical Specification limits are very infrequent, and are typically of short time duration, this approach of limiting dose rates l equivalent to the annual dose limits then assures that 10CFR20.106 limits on an annual average air concentration in unrestricted areas will be met.

i Each of tt.) methods to calculate dose or dose rate are presented in separate sections of Chapter 3, and are summarized in Table 1.1-1. Each Revision 6 - 2/18/88 ApprovedBy:4/[ [ w l

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5008R/12.332 3-3 l

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method has two levels of complexity and conservative margin and are called r Method I and Method II. Method I has the greatest margin and is the simplest; generally a linear equation. Method II is a more detailed analysis which allows for use of site-specific factors and variable parameters to be selected to best fit the actual release. Guidance is provided, but the appropriate margin and depth of analysis are determined in each instance at the time of analysis under Method II.

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3.2 Method to Calculate the Total Body Dose from Liquid Releases Technical Specification 3.11.1.2 limits the total body dose commitment J to a Member of the Public from radioactive material in liquid effluents to 1.5 mrem per quarter and 3 mrem per year. Technical Specification 3.11.1.3 requires liquid radwaste treatment when the total body dose estimate exceeds 0.06 mrem in any 31-day period. Technical Specification 3.11.4 limits the total body dose commitment to any real member of the public from all station sources (including 11gulds) to 25 mrem in a year. Oose evaluation is required at least once per 31 days. If the 11guld radwaste treatment system is not being used, dose evaluation is required before each release.

I Use Method I first to calculate the maximum total body dose from a liquid release from the plant.

Use Method II if a more accurate calculation of Total Body Dose is needed (i.e., Method I indicates the dose is greater than the limit), or if Method I cannot be applied.

I To evaluate total body dose for Specification 3.11.1.3 add the Total Body Dose from today's expected releases to the Total Body Dose accumulated l i for the time period of interest.

3.2.1 Method I The total body dose from a liquid release is:

D DFl (Eq. 3-1) tb " K f Oi itb (mrem) where:

OFLitb - Site-specific total body dose factor (mrem /C1) for liquid I release. See Table 1.1-7. l 01 - Total activity (Curies) released to liquids of radionuclide "1" during period of interest. For 1 Fe55, Sr89, Sr90, i or H3, use the best estimates (such as the most recent measurements).

Revision 6 - 2/18/88 g#h Approved By: #,/

5008R/12.332 3-5

K - 366/Fd ; where Fd is the average (typically monthly average) dilution flow of the Deerfield River below Sherman Dam (in ft3/sec). If Fd cannot be obtained or Fd IS greater than 366, K can be assumed to equal 1.0. The value 366 is the ten-year minimum monthly average Deerfield River flow rate below Sherman Dam (in ft3 /sec),

f Equation 3-1 can be applied under the following conditions (otherwise, justify I Method I or consider Method II):

1. Liquid releases to the circulating water pathway to Sherman l l Reservoir, or to the west storm drain pathway to the Deerfield River, and l

2. Any continuous or batch release over any time period.

l 3.2.2 Method 11 If Method I cannot be applied, or if the Method I dose exceeds the 1 limit or if a more exact calculation is required, then Method II should be applied. Method 11 consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific models, data or assumptions are more applicable. The base case analysis, documented below, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis.

I 3.2.3 Basis for Method I This section serves three purposes: (1) to document that Method I complies with appropriate NRC regulations, (2) to pr ovide background and training information to Method I users, and (3) to provide an introductory user's guide to Method II.

Method I may be used to show that the Technical Specifications which limit off-site total body dose from liquids (3.11.1.2, 3.11.1.3, and 3.11.4) have been met for releases over the appropriate periods. These Technical Specifications are based on design objectives and standards in 10CFR50 and Revision 6 - 2/18/88 Approved By:4 >

5008R/12.332 3-6 1

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40CFR190. Technical Specification 3.11.1.2 is based on the ALARA design objectives in 10CFR50, Appendix I Subsection II A. Technical Specification 3.11.1.3 is an "appropriate fraction", determined by the NRC, of that design objective (hereafter called the Objective). Technical Specification 3.11.4 is based on Environmental Standards for Uranium Fuel l Cycle in 40CFR190 (hereafter called the Standard) which applies to direct radiation as well as liquid and gaseous effluents. Method I applies only to the liquid contribution.

Exceeding the Objective or the Standard does not immediately limit l plant operation but requires a report to the NRC within 30 days. In addition, a waiver may be required. This is unlike exceeding 10CFR20 limits which could result in plant shutdown.

I Method I was developed such that "the actual exposure of an individual ... is unlikely to be substantially underestimated", (10CFR50, Appendix I). The definition, below, of a single "critical receptor" (a hypotheticci individual whose behavior iesults in an unrealistically high dose) provides part of the conservative margin to the calculation of total l body dose in Method I. Method 11 allows that actual individuals, with real behaviors, be taken into account for any given release. In fact, Method I was based on a Method II analysis for the critical receptor and annual average conditions instead of any real individual. That analysis was called the "base I case"; it was then reduceo to forrr Method I. The base case, the method of reduction, and the assumptions and data used are presented below.

The steps perfortred in the Method I derivation follow. First, in the base _caso, the dose impact to the critical receptor (in the form of dose factors in mrem /C1) for a 1 curie release of each radionuclide in liquid l effluents was derived. The base case analysis uses the methods, data and assumptions in Regulatory Guide 1.109 (Equations A-3, A-7, A-13 and A-16, l Reference A). Tables 3.2-1 and 3.2-2 outline human consumption and l environmental parameters used in the analysis. It is assumed that the critical receptor fishes below Sherman Dam and eats the fish caught from this locatien and consumes leafy vegetables and produce from a farm which is irrigated with water from the Deerfield River below Sherman Dam. Is is also Revision 6 - 2/18/88 Approved By: Jgf e g/ l wl/4 5008R/12.332 3-7

]

" assumed that the critical receptor drinks milk and eats meat from cows who

- drink water from the Deerfield River below Sherman Dam and eat silage from the irrigated farm above. The model is conservative because no real individual is likely to have that critical combination of exposures. A real individual could have only one or two pathways of exposure. A plant discharge flow rate of 308 ft /sec 3

was used with a mixing ratio of 0.84.

For any liquid release, during any period, the increment in annual average total body dose from radionuclide "i" is:

1 ADitb - 03 DFl itb where DFl itb is the total body dose factor for radionuclide "1" and Qg is the activity of radionuclide "1" released in curies.

Method I is more conservative than Method II because it is based on the following reduction of the base case. The dose factors, DFLitb, used in l Method I were chosen from the base case to be the highest of the four age l groups for + hat radionuclide. In effect, each radionuclide is conservativel)

I represented by its own critical age group.

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W W W W ( I U Table 3.2-1 Environmental Parameters for Liquid Effluents at Yankee Rowe (Derived from Reference A)

POTABLE AQUATIC SHORELINE FOOD GROWN WITH CONTAMINATED WATER VARIABLE WATER FOOD ACTIVITY VEGETABLES LEAFY VEG. MEAT COW MILK GOAT MILK

• MP Mixing Ratio 0.84 0.84 0.84 0.84 0.84 0.84 0.84 TP Transit Time (HRS) 12.0 24.0 0.0 0.0 0.0 480.0 48.0 48.0 YV Agricultural (KG/M2 ) 2.0 2.0 2.0 2.0 2.0 Productivity P Soll Surface (KG/M2 ) 240.0 240.0 240.0 240.0 240.0 Density IRR Irrigation Rate (L/M2/HR) 0.15 0.15 0.15 0.15 0.15 TE Crop Exposure (HRS) 1440.0 1440.0 1440.0 1440.0 1440.0 Time TH Holdup Time (HRS) 1440.0 24.0 2160.0 2160.0 2160.0 OAW Water Uptake Rate (L/D) 50.0 60.0 8.0 for Animal QF Feed Uptake Rate (KG/D) 50.0 50.0 6.0 for Animal Location of None Below Below Below Below Below Below None Critical individual Sherman Sherman Sherman Sherman Sherman Sherman Dam Dam Dam Dam Dam Dam

• Pathway is not included in Method I. It is listed for information purposes and the possible use in a Method II. l Revision 6 - 2/ N188 Approved By:,/f-[b ! ,Q, 500RR/17.332 3-9

Table 3.2-2 B e Specific Usage Factors for Various Liquid Pathways at Rowe (From Reference A, Table E-5. Zero where no pathway exists)

FISH INVERT. POTABLE SH0;i.

AGE VEG. LEAFY HILK MEAT WATER 1 (KG/YR)

VEG.

(KG/YR) (LITER /YR) (KG/YR) (KG/YR) (KG/YR) (LITER /YR) (HR.!

Adult 520.00 64.00 310.00 110.00 21.00 0.00 0.00 12 . -

Teen 630.00 42.00 400.00 65.00 16.00 0.00 0.00 67.:

Child 520.00 26.00 330.00 41.00 6.90 0.00 0.00 1 4 . '.

In7 ant 0.00 0.00 330.00 0.00 0.00 0.00 0.00 0. ( :

1 I

l 1

l l

I I

I I

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3.3 Method to Calculate Maximum Organ Dose from Liauid Releases Technical Specification 3.11.1.2 limits the maximum organ dose comaitment to a Member of the Public from radioactive material in liquid effluents to 5 mrem per quarter and 10 mrem per year. Technical Specification 3.11.l.3 requires liquid radwaste treatment when the maximum organ dose estimate exceeds 0.2 mrem in any 31-day period. Technical Specification 3.11.4 limits the maximum organ dose commitment to any real member of the public from all station sources (including liquids) to 25 mrem in a year except for the thyroid, which is limited to 75 mrem in a year. Dose evaluation is required at least once per 31 days. If the liquid radwaste treatment system is not being used, dose evaluation is required before each I rel(ase.

Use Method I first to calculate the maximum organ dose from a liquid release from the plant. I Use Method II if a more accurate calculation of organ dose is needed (i.e., Method I indicates the dose is greater than the limit), or if Method I cannot be applied.

I To evaluate the maximum organ dose for Specification 3.11.1.3, add the organ dose from the expected releases to the organ dose accumulated for the time period of interest.

3.3.1 Method I The maximum organ dose from a liquid release is:

D DFL gg (Eq. 3-2) organ (mrem)

Kf0 1 where:

DFLjm - Site-specific maximum organ dose factor (mrem /C1) for a liquid release. See Table 1.1-7.

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_ - _ _ _ _ _ _ _ _ _ _ _ _ _ . __ )

01

= Total activity (Curies) released to liquids of radionuclide "1" during period of interest. For 1 - Fe55, Sr89, Sr90, or H3, use the best estimates (such as the most recent measurements).

K - 366/Fd ; where Fd is the average (typically monthly average) dilution flow of the Deerfield River below Sherman Dam (in ft3/sec). If Fd cannot be obtained or Fd 15 greater than 366, K can be assumed to equal 1.0. The value 366 is the ten-year minimum monthly average Deerfield River flow rate below Sherman Dam (in ft3 /sec).

Equation 3-2 can be applied under the following conditions (otherwise, justify Method I or consider Method II):

1. Liquid releases to the circulating water pathway to Sherman l Reservoir, or to the west storm drain pathway to the Deerfield River, and

2. Any continuous or batch release over any time period.

3.3.2 Method II If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method II should be applied. Method 11 consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific models, data or assumptions are more applicable. The base case analysis, documented below, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis.

3.3.3 Basis for Method I This section serves three purposes: (1) to document that Method I complies with appropriate NRC regulations, (2) to provide background and

{ training information to Method I users, and (3) to provide an introductory user's guide to Method II. The methods to calculate maximum organ dose parallel the total body dose methods (see Section 3.2.3). Only the differences are presented here.

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( _ --- - - - - -

g For any liquid release, during any period, the increment in annual I average dose from radionuclide "i" to the maximum organ is:

ADgg Og DFL g l

l where DFL gg is the maximum organ dose factor for radionuclide "1" and Qg is the activity of radionuclide "i" released in curies.

The dose factors DFL j used in Method I were chosen from the base case to be the highest of the set of seven organs and four age groups for Pach radionuclide. This means that the maximum effect of each radionuclide is conservatively represented by its own critical age group and critical organ.

l l

l l

l l

I I

I I

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3.4 Method to Calculate the Total Body Dose Rate From Noble Gases l

Technical Specification 3.11.2.1 limits the dose rate at any time to the total body from noble gases at any location at or beyond the site boundary equal to or less than 500 mrem / year. By limiting the maximum Otb to a rate j I equivalent to no more than 500 mrem / year, assurance is provided that the total body dose accrued in any one year by any member of the general public will be less than 500 mrem in accordance with the annual dose limits of 10CFR Part 20 to unrestricted areas.

Use Method I first to calculate the total body dose rate from the peak release rate via the plant vent stack. Method I applies at all release rates.

Use Method I if M2thod I predicts a dose rate greater than the Technical Specification limit (i.e., use of actual meteorology over the period of interest) to determine if, in fact, Technical Soecification 3.11.2.1 had l actually been exceeded during a short time interval. I i

Compliance with the dose rate limits for noble gases is continuously l demonstrated when effluent release rates are below the plant vent stack noble gas activity rnonitor alarm setpctnt by virtue of the fact that the alarm l setpoint is based on a value wnich corresponds to the off-site dose rate limit I l

of Technical Specification 3.11.2.1, or a value below it. l Determinations of dose rates for compliance with Technical Specification (3.11.2.1) are performed when the effluent monitor alar-setpoint is exceeded and the corrective action required by Specification 3.11.2.1 is unsuccessful, or as required by the action to Technical l Specification Table 3.3.9 when the stack noble gas monitor is inoperable. i 3.4.1 Method I l

I The Total Body Dose Rate due to noble gases can be determined as follows: i I

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b( ) = 7.83 h gDFB g (Eq. 3.3) tb where:

Og

- In the case of noble gases, the telease rate from the plant vent stack (pC1/sec) of radionuclide "1". The release rate at the plant vent stack is based on the measured radionuclide distribution in the off-gas during plant operation, and the recorded total gas effluent count rate from the stack noble gas activity monitor. The release rate at the stack can also be stated in the following equation:

p h-d HhF g

'9 (Eq. 3-10)

$=( ) (cpm) (pCi/cc) cpm c

( sec c_c_ )

see where:

H = Plant vent stack monitor comt rate (cpm).

5 9

- Appropriate plant vent stack monitor detector counting efficiency (cpm /(pC1/cc)).

F = Plant vent stack flow rt:u (cc/sec).

dg - Fraction of the release which is nuclide "i". This fraction g can be based on the last measured value of nuclide "1" with respect to the total noble gas activity released at the PVS.

It can also be based on the fraction of nuclide "i" in the primary coolant wi.r. -espect to the total noble gas primary coolant activity.

OFBt - total body dose factor (see Table 1.1-2)

During periods (beyond the first two days) when the pla"t is shutddn

.6 ni radioactivity release rates can be measured at the PVS, Xo 133 may be

,i45 <l the referenced radionuclide to determine off-site dose rate and mon.sor setpoints. Alternately, a relative radionuclide "1" mix fraction

{ (f )g may be taken from Table 5.2-2 as a function of time after shutdown, and substituted in place of d g in Equation (3-10) above to determine the relative fraction of each noble gas potentially available for release to the

[ Approved By:

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i to' 1 Just prior to plant startup, the monitor alarm setpoints should be based on Xe-138 as reeresenting the most prevalent high dose factor noble gas expecied to be present shortly after the plant returns to power. Monitor I alarm setpoints which have been determined to be conservative under any plant conditions may be utilized a: any tirre in lieu of the above assumptions.

Equation 3-3 caa be applied under the follNing conditions (otherwise, 1 justify Method I or consider Method II):

1. Normal operath,as (not emergency event),

2. Noble gas releases via the plant vent stack to the atmosphere. l 3.4.2 Method II Tf Method I cannot be applied, or if the Method I dose exceous the lir'it oi if a more exact calculatio7 is reovired, then Method II may be g apolled. Method 11 consists of the models, input d;ta and assumptic in Regulatory Gu We 1.i,9, Rev. 1 (Reference A), except where site-spec) Tic models, data, or assumpt'ons are more applicable. The base case analysis, documented belcw, is a good example of the use of Method !!. It is an acceptable starting point for a Method II analysis.

3.4.3 Ba.is for Method _I.

This section serves four purposes: (1) to document that He'; hod I complies with appropriate NRC regulations, (2) to define the word "raie" as used in the Technical Spettftcation, (3) to provide backgrouni and training information to Method I users, and (4) to provide an introductory user's guide to Method !!.

Method I may be usec '

show that the Technical Specification which Ilmits total body dose rat < ole ga..t 'eleased to the atmosphere (Technical Specification 3.  ; has bt ' 'or the peak noble gas release rate.

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j

Technical Specification 1.11.2 l ensures "that the doses ... at and beyond the SITE BOUNDARY from gaseous effluents ... Will be within the annual dose limits ... associated with the concentrations of 10CFR Part 20, Appendix B, Table II, Column 1" (Technical Specification 3/4.ll.2.1-Basis). Those "Maximum Permissible Concentrations" for air in unrestricted areas, called H e n t.e MPC[0 , cannot be exceeded if this Technical Specification is met.

the requirements of 10CFR20.106(d) are met. Because the plant has never approachedevenasmallfractionofHPCfa limits Technical Specificatien 3.11.2.1 was given greater conservative margin by the NRC. It additionally restricts release rate monitor readings to the level at which the plant coulJ operate continuously and not exceed the annual dose limit. The annual total body dose limit is 500 mrem (from L35 Hardbook 69. Reference G Page 6), which limits.

isthebasilfortheHPC[A Exceeding the annual average total body dose rate could result in plant shutdown, especially if tne operators cannot take action to reduce the peak release rate.

Method I was derived f rom Regulatory Guide 1.109 as follows:

T 4 O 3.17 x 10 X/Q $ 7 {Q g0FB i

was derived by combining Equations B-4 and B-5 from Regulatory Guice i.109, assuming X/Q X/QU for ncble 94.45, and some simplification in the notation.

T Assuming that Df nite " '" "

tb " inite

.Q(pC1/sec)*31.54/Q(C1/yr) we get: /

b (mremlyr) tb

- 1 x 106 5 (X/Q)Y I Qg 0FB g i

substitu'ing:

S F

- 1.0 (shielding factor)

[X/Q)Y Long-Term Average Gamma Dilution factor

- 7.83 x 10-6 (3,cj,3)

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-- _- - - _ _ _ _ _ _ _ _ _ 1

hg - Release rate of noble gas "1" (pC1/sec) gives:

( . .

Otb(mrem /t.)=7.83IQg(pC1/sec)DFBg (Eq. 3-3) 1 MethodIIcannotprovidemuchextrarealismbecauseb tb is already based on several factors which make use of current plant parameters. However, should it be needed, the dose rate analysis for critical receptor can be performed making use of current meteorology during the time interval of H recorded peak release rate in place of the default atmospheric dispersion factor used in Method I.

t L

L

[

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3.5 Mtthod to Calculate the Skin Dose Rate from Noble Gases Technical Specification 3.11.2.1 limits the dose rate to the skin from noble ga'.es at any location at or beyond the site boundary to 3,000 mrem / year.

l .

By limiting the maximum Dsk to a rate equivalent to no mor9 than 3,000 mrem / year, assurance is provided that the skin dose accrued in any one year by any member of the general public is much less than 3,000 mrem.

Use Method I first to calculate the skin dose rate from the peak relcase rate via the plant vent stack. Method I applies at all release rates. ,

Use Method II if Method I predicts a dote rate greater than the Technical Specification limits (i.e., use of actual meteoro' gy over the period of interest) to determine if, in fact, Technical Specification 3.8.E.1 had actually been ex*'eeded during a short time interval.

I Compliance with the dose rate limits for noble gases is continuously demonstrated when effluent release rates are below the plant vent stack noble gas activity monitor alarm setpoint by virtue of the fact that the alarm  !

setpoint is baseri on a value which corresponds to the off-site Technical Specification dose rate limit, or a value belon it. .

Determinations of dose rate for compliarce with Technical Specification (3.11.2.1) are performed when the efflue.it monitor alarm setpoint is exceeded and the corrective action required by Specification l t 3.11.2.1 is unsuccessful, or as required by the notations to Technical Specification Table 3.3.9 when the stack noble gas monitor is inoperable, j

=

3.5.1 Method I )

The Skin Dose Rate due to noble gases 's:

(

bskin (*#'*#Y " I i 0 E9' 3-D

, . /

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L where:

Qg - n the case of noble gases, the release rate from the plant l vent stack (pC1/sec) of each radionuclide "1". The release rate at the plant vent stack is based on the measured radionuclide distribution in the off-gas during plant operation, and the recorded total gas effluent count rate from ,

the stack noble gas activity monitor. The release rate at the  ;

stick can also be stated in the following equation:

hd Hff g 9 '

I (Eq. 3-10) '

h-( ) (cpm) (u cc) ( )

where:

f 1

H = Plant vent stack monitor count rate Mpm).

Sg - Appropriate plant vent stack monitor detector counting efficiency (cpn/(pC1/cc)).

F - Plant vent stack flow rate (cc/ set). .

dj - Fraction of the release which is nuclide "i". Thi; fraction can be based on the last measured value of nuclide "i" with  ;

respect to the total noble gas activity released at the PVS, It can also be based on the fraction of nuclide "i" with l respect to the total noble gas primcry coolant activity.

i DFj - skin dote factor (see Table 1.1-2)

During periods (beyond the first two days) when the plant is shutdown and no radioactivity release rates can be measured at the PVS, Xe-133 may be used as the referenced radionuclide to d?termine off-site dose rate and j monitor setpoints. Alternately, c triative radionuclide "1" mix fraction (f )g may be taken from Table 5.2-2 as a function of time after shutdown, and substituted in place of d in g Equation (3-10) above to determine the ,

relative fraction of each noble gas potentially available for release to the total. Ju'st prior to plant startup, the monitor alarm setpoints should be based on Xe-138'as representing the most prevalent high dose factor novle gas expected to be present shortly after the plant returns to power. Monitor Revision 6 - Zi M! G Approved By: .4-fg f 7 4 b

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alarm setpoints which htve been determined to be conservative under any plant conditions may be utilized at any time in lieu of the above assumptions.

Equation 3-4 e.an be applied under the fo? lowing conditions (otherwise, justify Method I or consider Method II):

1. Normal operations (not emergency event),

2. Noble gas releases via the plant vent stack to the atmosphere. l 3.5.2 Method II If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method II may be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109 Rev. 1 (Reference A), except where site-specific models, data or assumptions are more applicable. The base case analysis, documented below, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis.

3.5.3 Basis For Method I This section serves four purposes: (1) to document that Method I complies with appropriate NRC regulations, (2) to define the word "rate" as used in the Technical Specification, (3) to provide background and training information to Method I users, and (4) to provide an introductory user's guide to Method II. The methods to calculate skin dose rate parallel the total body

( dose rate methods in Section 3.4.3. Only the differences are presented her_e.

Method I may be used to show that the Technical Specification which

{ limits skin dose rate from noble gases released to the atmsphere (Technical Specification 3.11.2.1) has been met for the peak noble gas release rate.

The annual skin dose limit is 3,000 mrem (from NBS Handbook 69, Reference G Pages 5 and 6, is 30 rem /16, which is the basis for the HPC["Iim)t'..

Revision 6 - 2/18/88 Approved By: Afb 5008R/12.332 3-21

l Method I was derived from Regulatory Guide 1.109 as follows:

S D 3.17 x 104 (X/Q 1.11 S F Ii 0 DF{

1 + X/Q {i Q DFS g g) eas derived by combining Equations B-4, B-5, and B-7 from Regulatory Guide I 1.109, assuming that X/Q - X/Q for noble gases, and making some simpitfications in notation. Assuming that Dfinite " 0 #0 # #03 and that Osk - DS

• Q(pCi/sec) = 31.54/Q(C1/yr) yields bskin (mrem /yr) - 1.11 x S F x (1 x 10b (X/Q)Y h DF{

+1x 106 (X/Q) I h DFS g i

where:

3 I

[X/Q)Y 7.83 x 10-0 sec/m i

= 2.39 x 10-5 3,cfm 3 l X/Q Sp 1.0 (shielding factor) l Subs'.ituting gives f I

l 6,un (nre"yr) 8 69 I 6, Dr7 + 23.9 x 10 { h DFS g g i 1 I1 h (8.69 g DF{ + 23.9 DFSg )

Define:

DFg-8.69DF{+23.9DFS g Then:

O skin I"'* /Y ") " I 0 DFg1

,2q. 3-4) 1

/ //

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s

( 3.6 Method to Calculate the Critical Organ Dose Rate from 131-I, 3-H and Particulates with T 1/2 Greater Than 8 Days Technical Specification 3.11.2.1 timits the dose rate to any organ from 131-1, 3-H and radionuclides in particulate form with half lives greater than 8 days to 1500 mrem / year to any organ.

The peak release rate averaging time in the case of todines and

( particulates is commensurate with the time the lodine and particulate samplers are in service between changeouts (typically a week). By limiting the maximum bg ,to a rate equivalent to no more than 1500 mrem / year, assurance is provided that the critical organ dose accrued in any one year by any member of the general public will be less than 1500 mrem.

Use Method I first to calculate the critical organ dose rate from the peak release rate via the plant vent stack. Method I applies at all release  ;

rates. l i

Use Method II if Method I predicts a dose rate greater than the Technical Specification limits (i.e., use of actual meteorology over the period of interest) to determine if, in fact, Tech 11 cal Specification 3.11.2.1 had actually been exceeded during the sampling period. ,

3.6.1 Method I The critical organ dose rate can be determined as follows:

b,-

e hg 0FGje, (Eq. 3-5) crem gQt mrem-see yr sec pCl-yr

[

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where:

- I g Og - Stack activity release rate determination of radionuclide l "1" (todine, tritium, and particulates with half-lives l greater than 8 days) in pC1/sec. For i - Sr89, Sr90, or H3, use the best estimat?s (such as most recent measurements). I

'*~5'C Site-specific critical organ dose rate factor (*'Ci-yr ) for I DFG'CO I

a gaseous release. (See Table 1.1-8.)

p Equation 3-5 can be applied under the following conditions (otherwise, l justify Method I or consider Method II):

1 1. Normal operations (not emergency event),

2. Tritium, iodine, and particulate releasei via the plant vent l stack to the atmosphere As an alternative to performing Method I calculations, compliance with the critical organ dose rate limit of 1,500 mrem / year in Technical Specification 3.ll.2.1.b can be shown by two methods. They are: a comparison of the measured I-131 release rate to determine if it is at or below an inspection limit of 0.0125 pC1/sec, or a concentration limit in the PVS equivalent to:

C b

I-131 (pCi/cc) . (2.65 x 10-5)/F (cfm) where:

F = average PV5 flow rate measured during the sampling interval.

This results from the fact that I-131 is the controlling nuclide with respect to any critical organ dose, and the selected inspection limits represent approximately ten percent (10 percent) of the 1,500 mrem / year dose rate I limit. Measured values greater than the inspection limits should be evaluated y equation 3-5, or a Method !! assessment.

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-- __ __ _ }

3.6.2 Method !!

If Method I cannot be applied, or if the Method I dose exceeds the Technical Specification limit or if a more exact calculation is required, then Method !! may be applied. Method 11 consists of the models, input data and assumptions in Regulatory Guide 1.109, Revision 1 (Reference A), except where site-specific models, data or assumptions are more applicable. The base case analysis, documented below, is a good example of the use of He' hod !!. It is an acceptable starting point for a Method II analysis.

3.6.3 Basis for Method I This section serves four purposes: (1) to document that Method I I compiles with appropriate NRC regulations (2) to define the word "rate" as used in the Technical Specification, (3) to provide background and training information to Method I users, and (4) to provide an introductory user's guide to Method II. The methods to calculate critical organ dose rate parallel the total body dose rate methods in Section 3.4.3. Only the dif_ferences are l presented here.

l Method I may be used to show that the Technical Specification which limits organ dose rate from lodine-131, Tritium and radionuclides in g

5 part** alate form with half lives greater than 8 days released to the l atmosphere (Technical Specification 3.11.2.1) hhs been met for the peak >

Iodine-131, tritium, and particulate release rates.

I Theequationforb e

is derived by modifying Equation 3-8 from Section 3.9 as follows:

i l

D gg -

IQg DfG ggo i

(Eq. 3-8) mrem O "*

yr Ci applyirig the conversion f actor, 31.54 (Ci-sec/pCl-yr) andconvertingQtohin pC1/sec yield; -

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5008R/12.332 3-25

D - 31.54 I Q DFGgCg g

eo '

F

' C1-sec gCi mrem mrem yr pCi-yr sec Ci

[ Eq. 3-5 is rewritten in the form:

I U

co I0 1 DFGjgo 1

L where:

r CFGjgo - DFG gco 31.54 mrem-see mrem Ci-sec pCi-yr Ci pCi-yr Should Method II be needed, the analysis for critical receptor critical pathway (s) and the annual average dispersion coefficients may be performed with actual meteorologic and latest land use census dita to ider.tify the location of those pathways which are most impacted by these types of releases.

~

I I i I

I ,

I l

l l

Revision 6 - 2/18/88 Approvec By:AI[ j 5008R/12.332 3-26

I 3.7 Method to Calculate the Gama Air Oose from Noble Gases u Technical Specification 3.11.2.2 limits the gamma dose to air from p roble gases at any location at or beyond the site botndary to 5 mrad in any quarter and 10 mrad in any year. Dose evaluation is required at least once per 31 days.

Use Method I first to calculate the gamma air dose for the plant vent stack releases during the period. Method I applies at all dose levels.

Use Method 11 if a more accurate calculation is needed, or if Method I cannot be applied.

, 3.7.1 Method I The gam a air dose from plant vent stack releases is:

(Eq. 3-6)

D tr(mrad) = 0.25 ' IQ 0F{

3 l where: '

Qg - total activity (Curles) released to the atmosphere via the plant I vent stack of each radionuclide "i" during the period of interest.

DF{

gama dose factor to air for radionuclide "1". See Table i.1-2 Equation 3-6 can be applied under the following conditions (otherwise justify '4ethod I or consider Method II):

1. Normal operations (not emergency eunt),

2. Noble gas releases via the plant vent stack to the atmosphere.

I I

, , /

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r 3.7.1.1 Ground !.evel Releases For ground level releases, the gamma air dose is:

1 Uq. 3-6.1)

Ogrd (mrad) - (1.23 x 10-4) (QXe-133 equivalent) where:

OXe-133 equivalent - the Xe-133 equivalent of all the noble gases in the release (curies), and is based on the dose conversionfactors(DF{}aslistedin Table 1.1-2 of the ODCM.

3.7.2 Method II If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method II may be I applied. Method 11 consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev 1 (Reference A), except where site-specific i models, data or assumptions are nore applicable. The base case analysis, dccumented below, is a good example of the use of Method 11. It is an acceptable starting point for a Method II analysis.

3.7.3 Basis for Method I This section serves three purposes: (1) to document that Method I Complies with appropriate NRC regulations, (2) to provide background and training information to Method I users, and (3) to provide an introductory I user's guide to Method II.

Method I may be used to show that the Technical Specification which limits off-site gamma air dose from gaseous effluents (3.11.2.2) has been met for releases over appropriate periods. This Technical Specification is based on the Objective in 10CrR50, Appendix I, Subsection B.1, which limits the estimated annual gamma air dose at unrestr'.cted area locations.

/ /

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L Exceeding the Objective does not immediately limit plant operation but

~

requires a report to the NRC within 30 days.

I For any noble gas release, in any period, the dose is taken from '

Equations B-4 and B-5 of Regulatory Guide 1.109 with the added assumption that Dfinite = 0 (X/Q) M X/Q):

I 3

}

Dair(mrad) 3.17 x 104(h['-) (X/Q)Y (sec/m ) Q(C1)DF{( g rp ,

where:

[X/0)Y = long-term average gamma dilution factor I 7.83 x 10"' (sec/m )

3 Qg

- num'ser of curies of noble gas "1" released which leads to:

(Eq. 3-6) 0 tr(mrad) 0.25 { iQ (C1) g DF}

l The main difference between Method I and Method II i, that Method 11  !

would allow the use of actual meteorology to determine (X/Q)Y rather than I use the maximum long-term average value obtained for the time period from 1/81 through 12/S5.

i The gamma-air dose from a ground level release is determined by Jsing the sane Regulatory Guide 1.109 equation to derive Equation 3-6 above. The only differences are:

1 (X/Q)Y - 1.1 x 10-5 g,gj,3 , which is the long-term average grourid level (X/Q)Y based en the time period from I May 1977 through April 1982.

O xe-133 equivalent = the Xe-133 equivalent of all the noole gases in the release (curies).

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L

• 3 DF} = DF{,,j33 3.53 x 10'4 ( }' ) obtained from Table 1.1.2. to L

to account for the release being expressed in terms of r Xe-133 equivalent. ,.

L ,

Substituting the above into the Regulatory Guide 1.109 general equation

[ O grd (mrad) = 1.23 x 10~4 Q Xe-M3 equivalent. (Eq. 3-6.1)

[

[

[

[

[

[

[

[

[

[

[

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[.

f _ - - - - - - - - - - -

L 3.8 Method to Cal:ulate the Beta Air Dose from Noble Gases Technical Specification 3.11.2.2 limits the beta dose to air from noble r gases at any location at or beyond the site boundary to 10 mrad in any quarter ard 20 mrad in any year. Dose evaluation is required at least once per 31 days.

Use Method I first to calculate the beta air dose for the plant vent

( stack releases during the period. Method I applies at all dose levels.

[ Use Method II if a move accurate calculation is needed of if Method I cannot be applied.

[ 3.8.1 Method I The beta air dose from plant vent stack releases it:

(Eq. 3-7)

D r(mrad) = 0.76 iI Q g0F

[ where:

See Table 1.1-2 Off beta dose factor to air for radionuclide "t".

{

Qg total activity (Curles) released to the atmosphere via the plant

( vent stack of each radionuclide "1" during the period of interest.

Equation 3-7 can be applied under the following conditions (otherwise

{ justify Method I or consider Method II):

.b 1. Normal operations (not emergency event), and

2. Noble gas releases via the plant vent stack to the atmosphere.

[

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3.8.1.1 Ground Level Releases For ground level release',, the beta air dose can be determined by using 1 Equation 3-7. Equation 3-7 results in doses that are approximately 10 percent more conservative than calculating releases using ground level methodology.

3.8.2 Method I.I.

If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method II may be applied. Method !! consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific .

models, data or assumptions are more applicable. The ba u case analysis, docu.?ented below, is a good exartple of the use of Method II. It is an I acceptable starting point for a Method II analysis.

3.8.3 Basis for Method I This section serves three purposes: (1) to document that Method I complies with appropriate NRC regulations, (2) to provide background and training information to Method I users, and (3) to provide an introductory user's guide to Method II. The methods to calculate beta air dose parallel the gamma air dose methods in Section 3.7.3. Only the differences 3r2 I presented here.

Method I may be used to show that the Technical Specification which limits off-site beta air dose from gaseous effluents (3.11.2.2) has been tret for releases over appropriate periods. This Technical Specification is based on the Objective in 10CFR50, App (ndix 1. Subsection B.1, which limits the estimated annual beta air dose at unrestricted area locations.

Exceeding the Objective does not immediately limit plant operation but requires a "eport to the NRC within 30 days.

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T For any noble gas release, in any period, the dose is taken from F Equations B-4 and B-S of Regulatory Guide 1.109:

L ,

{ D tr(mrad) = 3.17 x 104 (X/0) {0DFf 3 substituting X/Q = 2.39 x 10' sec/m3

[ He have (Eq. 3-7)

{ D tr(mrad) = 0.76 Q3 (C1) DFf E

[

[

[

d t

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3.9 Method to Calcul4te the Critical Organ Dose from Tritium, Iodines and Particulates l Technical Specification 3.11.2.3 limits the critical organ dose to a Member of the Public from radioactive Tritium, Iodine-131, and particulates with half-lives greater than 8 days in gaseous effluents to 7.5 mrem per quarter and 15 mrem per year. Technical Specification 3.11.4 limits the total body and organ dose to any real member of the public from all station sources (including gaseous effluents) to 25 mrem in a year except for the thyroid, which is limited to 75 mrem in a year.

Use Method I first to calculate the critical organ dose from a vent stack release as it is simpler to execute and more conservative than Method II. l Use Method II if a more accurate calculation of critical organ dose is needed (i.e., Method I indicates the dose is greater than the limit), or if Method I cannot be applied. l 3.9.1 Method I l

The critical organ dose fron a gaseous release is: l DFG ggo (Eq. 3-8) l U

co (mren)

"f0 1 where:

Qg - Total activity (Curies) released to the atmosphere of radionuclide i during the period of interest. For 1 -

Sr-89, Sr-90, or H-3, use the best estimates (such as the I most recent measurements).

0FGt co - Site-specific critical organ dose factor (mrem /C1) for a 1 gaseous release. See Table 1.1-8.

Equation 3-8 can be appiled under the following conditions (otherwise, justify Method I or consider Method II):

1. Normal operations (not emergency event),

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2. Iodine, tritium, and particulate releases via the plant vent stack l to the stmosphere, and

{

s 3. Any continuous or batch release over any time pertod.

3.9.2 HETH00 II If Method I cannot be applied, or if the Method I dose exceedt the limit or if a more exact calculation is required, then Method !! should be applied. Method II consists of the models, input data and assumptions in l Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific models, data or assumptions are more applicable. The base case analysis, documented below, is a good example of the use of Method !!. It is an acceptable starting point for a Method II analysis.

I 3.9.3 Basis for Method I This section serves three purposes: (1) te document that Method I complies with appropriate NRC regulations, (2) to provide background and training information to Method I users, and (3) to provide an introductory user's guide to Hethod II.

Method I may be used to show that the Technical Specifications which I limit off-site organ dose from gases (3.11.2.3 and 3.11.4) nave been met for releases over the appropriate periods. These Technical Specifications are based on Objectives and Standards in 10CFR and 40CFR, Technical Specification 3.11.2.3 15 ba'ed on the ALARA Objectives in 10CFR50, Appendix I, Subsection

!! C. Technical Specification 3.11.4 is based on Environmental Standards for Uranium fuel Cycle in 40CFR190 (hereafter called the Standard) which applies to direct radiation as well as Itquid and gaseous effluents. These methods apply only to iodine, tritium, and particulates in gaseous effluents coritribution.

I I

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Exceeding the Objective or the Standard does not immediately limit

[ plant operation but requires a report to the NRC within 30 days. In addition, a waiver may be required.

Method I was developed such that "the actual exposure of an I individual ... is unlikely to be substantially underestimated" (10CFR50, Appendix !). The use below of a single "critical receptor" provides part of the conservative margin to the calculation of critical organ dose in Method I. Method II allows that actual individuals, with real behaviors, be taken into account for any given release. In fact, Method I was based on a Hsthod 11 analysis of the critical receptor for the annual average conditions. For purposes of complying with the Technical Specifications 3.11.2.3 and 3.11.4 annual average dilution factors are appropriate for batch and continuous releases. That analysis was called the "base case"; it was then reduced to I form Method I. The base case, the method of reduction, and the assumptions and data used are presented below.

The steps performed in the Method I derivation follow. First, in the base case, the dose impact to the critical receptor in the form of dose factors, DrGgg,(trem/Ci) of I curie release of each iodine, tritium, and l particulate radionuclide to gaseous effluents was derived. Then Method I was determined using simplifying and further conservative assumptions. The base case ar-lysis uses the methods, data and assumptions in Regulatory Guide 1.109 I (Equations C-2, C-4 and C-13 in Reference A). Tables 3.9-1 and 3.9-2 outline human consumption and environmental parameters used in the analysis. It is conservatively assumed that the critical receptor lives at the "maximum site boundary dilution factor location" as defined in Section 3.10.

For any gas release, during any period, the dose from radionuclide "1" is:

Olco " 0 ico0 i where DFGgg,is the critical dose factor for radionuclide "i" and Qg is the activity of radionuclide "i" released in curies.

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l Method I is more conservative than Method II in the region of the p Technical Specification limits because it is based on the following reduction H of the base case. The dose factors DFG ggo used in Method I were chosen from e the base case to be the highest of the four age groups for that radionuclide.

L In effect, each radionuclide is conservatively represented by its own critical age group and critical organ.

[

E

[

[

[  !

[

[

[

[

[

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Tabla 3.9-2 Environmental Parameters for Gaseous Effluents at Yankee Rowe (Derived from Reference A)

Vegetables Cow Milk Goat Milk

• Meat l Variable Stored Leafy Pasture Stored Pasture Stored Pasture Stored YV Agricultural (Kg/M2 ) 2. 2. 0.7 2. 0.7 2. 0.7 2. l Productivity P Soil Surface (KGIM2 ) 240. 240. 240. 240. 240. 240. 240. 240.

Density ,

1 I T Transport Time (HRS) 48. 48. 48. 48. 480. 480.

to User TB Soi1 Exposure (HRS) 131400. 131400. 131400. 131400. 131400. 131400. 131400. 131400.

i Time TF Crop Exposure (HRS) 1440. 1440. 720. 1440. 720. 1440. 720. 1440. l Time to Plume TH Holdiep After (HRS) 1440. 24 O. 2160. O. 2160. O. 2160.

Harvest 0F Animals Daily (KG/ DAY) 50. 50. 6. 6. 50. 50.

Feed FP fraction of Year 0.50 0.50 0.50 on Pasture

• Pcthway is not included in Method I. It is listed for information purposes, and the possible use In a Method II analysis.

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M U ~l f L___I Tcb10 3.9-2 (continued)

Environmental Farameters for Gaseous Ef'luents at Yankee Rowe (Derived from Reference A)

Vegetables Cow MIIk Goat Milk

• Meat l f Pasture Stored Pasture Stored Pasture Stored Variable Stored Leafy l

1. 1. 1.

FS Fraction Pasture t: hen On Pasture FG Fraction of Stored 0.76 Veg. Grown in Garden FL Fraction of leafy 1.0 Veg. Grown in Garden FI fraction Elemental Iodine - 0.50 H Absolute (gm/M3 )

Humidity - 5.6

• Pathway is not included in Method I. It is listed for information purposes, and the possible use in a Method II analysis.

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L 3.10 Critical Receptors and Long-Term Average Atmospheric Dispersion Factors for Important Exposure Pathways The gaseous effluent dose eqJations (Hethod I) have been simplified by assuming an individual whose behavior and living habits inevitably lead to a higher dose than anyone else. The following pathways of exposure to gaseous effluents as listed in Regulatory Guide 1.109 (Reference A) have been considered. They are:

Direct exposure to contaminated air, Direct exposure to contaminated ground, Inhalation of air, Ingestion of vegetables, Ingestion of cow milk, and Ingestion of meat.

Section 3.10.1 details the sclection of important off-site locations and receptors; Section 3.10.2 describes the atnespheric model used to convert meteorologic data into dispersion factors; and Section 3.10.3 contains the

1 ting descriptions of the critical receptors and their dispersion factors as

( rr a function of exposure pathway.

3.10.1 Critical Receptors The most limiting site boundary location in which individuals are, or l Itkely to be, located was assumed to be the receptor for all the gaseous pathways considered. This provides a conservative estimate of the dose to an individual from existing and potential gaseous pathways for the Method I analysis.

This point is the SSE sector, 800 meters.

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3.10.2 Yankee Atmospheric Dispersion Model The annual average dispersion factors are computed for routine (long-term) releases using the Yankee Atomic Electric Company's (YAEC) AEOLUS '

( (Reference B) Computer Code.

AEOLUS produces the following annual average dispersion factors for each

(

location:

- EQ, nondepleted dispersion factors for evaluating ground level concentrations;

- (X/Q)D. depleted dispersion factors for evaluating ground level concentrations of iodines and particulates; Y

X_/Q , effective gamma dispersion factors for evaluating gamma

( -

dose rates from a sector-averaged finite cloud (multiple-energy undepleted source); and

{

- D/Q deDosition factors for dry deposition of elemental

[ radiolodines and other particulates.

( The AEOLUS diffusion model is described in the AEOLUS Computer Code Manual (Reference B). AEOLUS is based, in part, on the straight-1tne airflow model as discussed in Regulatory Guide 1.111 (Reference C).

(

One difference is that gamma dose rate is calculated throughout this

{ 00CM using the fintte cloud model presented in Meteorology and Atomic Energy 1968 (Reference H, Section 7-5.2.5). That model is implemented through the definition (Reference B, Section 6) of an effective gamma dispersion factor, X/QY , and the replacement of X/Q in infinite cloud dose equations by the

( X/QY .

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r --- - -

L f The other difference is that the relatively narrow valley in which the f plant sits is considered by the model. Wind channelling is assumed to occur I in the seven sectors which make up the valley. The seven sectors are SSE, S, H SSW, SW, WSW, W, and WNW. If a receptor location is in one of the valley L sectors, the contribution from the other six valley sectors are averaged into the particular valley receptor. This is done for distances greater than 500 meters from the primary vent stack where the valley effects are assumed to cause channelling.

3.10.3 Long-Term Average Dispersion Eactors for Critical Receptors Actual measured meteorological data for the five-year pericd, 1/81 through 12/85, were analyzed to determine the locations of the maximum off-site atmospheric dispersion factors. Each dose and dose rate calculation incorporates the maximum applicable off-site, long-term average atmospheric dispersion factor. The values used and their locations are summarized in j Table 3.10-1.  !

t '

E

[

[

[

[

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7 n W n n W - n n m TABLE 3.10-1 Yankee Nuclear Pcwer Station Five-Year Average Atmospheric Dispersion Factors

• Dose to Critical Dose Rate to Individual Dose to Air _ Organ Total Body _ Skin Critical Or_gan Gamma Beta Thyroid 2.19E-06 2.19E-06 l X/0dpleted(Sf) - -

m X/0undepleted(5f) - 2.39E-06 - -

2.39E-06 -

m 5.02E-08 - -

5.02E-08 0/0(h) m 7.83E-06 7.83E-06 -

7.83E-06 - -

X/0Y (503)

• SSE st'e boundary, 800 meters from Primary Vent Stack l

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3.11 Method to Calculate Direct Oose from Plant Operation Technical Specification 3.11.4 restricts the Dose to the whole body and any organ of any real member of the public from all station sources (including '

( direct radiation from the reactor and outside storage tanks which is called the Direct Dose) to the limit of 25 mrem ia a year, except for the thyroid which is limited to 75 mrem in a year. A determination of the need to conduct

{ a total dose evaluation is required at least every 31 days.

Use Method I first to calculate the Olrect Dose contribution to the whole body and any organ as it is simpler to execute and more conservative than Method II.

Use Method II if a more accurate calculation of Direct Dose is needed or if Method I cannot be applied.

3.11.1 Method I The maximum contribution of Direct Dose to the whole body or to any organ is:

O d-(0057+k)T,0.00087 r

(Eq. 3-9)

{

where:

0.00087 - Conversion t' actor (mrem /pR).

Te - Length of exposure period in hours.

E Exposure rate at critical receptor from non-V.C. Sources as r measured or estimated for the period.

Equation 3-9 can be applied under the following conditions, otherwise justify

( Method I or consider Method II:

1. Normal operations (not emergency event),

s /

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[

f - - - - - -

r

2. Allsignificantremainingsourcesareconsideredink,and r

3. Any exposure period.

3.3.2 Method II If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method II should be applied. Methoo II consists of measurement and site-specific models, data and assumptions. The base case analysts, documented below, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis.

3.3.4 Basis for Method I This section serves three purposes: (1) to document that Method I complies with appropriate NRC regulations, (2) to provide background and training information to Method I users, ar.d (3) to provide an introductory user's guide to Method II.

Methoa I may be used to show that the Technical Specification that limits Direct Dose off-site (3.11.4) has been met for any exposure period.

{ Technical Specification 3.11.4 is based on the Standard (40CFR190) which applies to direct sources of radiation as well as liquid and gaseous effluents. Method I applies to the direct sources only.

Exceeding the Standard does not immediately limit plant operation but requires a report to the NRC within 30 days. In addition, a walver may be required. This is unlike exceeding 10CFR20 limits which could result in plant shutdown.

Method I is developed below by reducing the "base case" (a Method II J analysis) using conservative assumptions. The base case involves the choice of a critical receptor and the development of an exposure factor for the vapor container source, EVC, (pR/ hour operation). The critical recepter is the l

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[ ,

nearest resident who lives 450m from the Vapor Container centerline in the NNW direction. An occupancy of 1.0 is assumed.

The exposure factot k is developed below by extrapolating VC measurements made close to the plant, out to the critical receptor. All I

significant sources of direct radiation on-site are shielded by buildings and tanks from the critical receptor with the exception of the Vapor Container and one of the liquid waste storage tanks, that is Number TK-31.

The dose (mrem) to the critical receptor, Ddover the exposure period l (in hours), T,, is related s kplistically to the exposure rate from the Vapor Container in pR/ hot.r EVC, and the exposure rate from remaining sources, Er , by the following equation:

1 D d" VC + r) T, 0.00087 (Eq. 3.1-1) l What remains is to conservatively derive:

k VC andk r The dote from the Vapor Container is due to fission end activation gases which build up in plant systems and the Vapor Container during operation. Those sources decay or are ventilated through the Ventilation Exhaust Filtration System at the beginning of refu ling outages.

This has to be done to allow worker access to the Vapor Container to reload the core. The estimate of E is based on the extrapolation of VC ,

I measurements made during plant operation at the restricted area fence, E),

compared to background measurements made during a refueling outage after containment purge at the same locations, E ' 5 expected to remain b VC constant over the years and so it can be estimated here as a function of exposure period.

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Although regular measurements of direct radiation are made near the I

critical receptor as part of the environmental surveillance program, the L,

majority of the measured coses are due to natural background and fallout, variations in which entirely obscure plant contributions to dose. Because ,

7 they are closer to the sources, the measurement of direct radiation using TLDs

) at the restricted area fent.e can be extrapolated with greater net sensit vity l (about 10 mR/ year). However, the most sensitive method proveu to be exposure rate measurements madt with a High Pressure Ionization Chamber, which had a history of ipr / hour plus or minus lpR/ hour 95 percent confifence interval for ]

exposure rates near the background rate for the procedure used (20 replicate measurements and periodic instrument checks). This extrapolates to approximately a plus or minus Imrem/ year 95 percent confidence interval at the r critical receptor, k is estimated using the following equation:

VC L

. d I

do . 3. l b 2 )

[ E VC " d

~

b) where:

1 d = Olstance to critical receptor from Vapcc Container centerline.

~

d, = N stance to exposure measurement from Vapor Container centerline.

k Exposure rate (pR/hr) measurement during plant operation, k Er.posure rate (pR/hr) measurement during plant outage.

b I

k is derivea from data collected in 1981 and presented in Table VC r 3.11-1. Tne mean value of E.,, for t..easurements at each of the nine TLD locations at the restricted ree fence is 0.057 pR/ hour. The ruan value is used because it is insensitive to miscellareoa on-site sourcos which contricute to the measurements but not to the dose at the critical receptor, krwill have to bt made from measurements or estimates made for the specific exposure period.

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u Substitutingthederivedvalueofk VC into Equation 3.11-1 yields:

b O d-(0.057+d)T,0.00087 (Eq. 3-9) r

[

[

o

[

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l

f Table 3.11-1 F Estimate of Exposure Rate at Critical Receptor From Vapor Container Shine, L EVC, Made in Spring 1981 Monitoring Station No. d Direction k k VC o b (km) (pR/hr) (pR/hr) (pR/hr) 13 .08 2250 20.4+ 18.0* .076 14 .11 3000 15.0 13.2* .108 15 .08 3450 14.4 13.1* .041 16 .13 300 17.4 16.6* .067 17 .14 700 15.9 15.6 " .029 18 .14 1150 2'.9 24.6 " .029 19 .16 1400 23.2 23.3 " .013 20 .16 1600 19.4 18.B" .076 18.7 '

[ 21 .11 2050 20.3 .096 Total - 0.509 Average - 0.057 pR/hr

+ All reasures of E taken 4/28/81, average daily power level was 97.5 HWe-Net, 5 days before shutdown at end of cycle 13/14.

• Heasures of k taken 5/13/81, during outage. Containment purged.

b

[ ' Heasures of d taken 6/2/81, during shutdown. Containment purged.

b f '

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500SR/12.332 3-50 f . _ - _ - - - - -

l l

I

,, 4 O 50 100 5'.

I' I , , , e f _l ,j i.k m

,..r.

.: .. , c.

f e..

METERS -

F^'^' 4

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Cate

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House . *

,- . . . . . . , N, l Switch-8 offaces

l yardg l CM.16 (ej

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@ gv,.13 - , Aug. gay l Area -m **

Center *

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Con t a t rie r t W 'i'#*, ,' 4 . . . . * ' *

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4 Figure 4-4 Yankee Plant Radiological Environe. ental Monitoring Locations at the Restricted Area Fence (Direct Radiation Pathway)

[ Revision 6 .'/16/65 4-5 Approved By:M', e/s/MnA.

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L Table 4-1 Radiological Environmental Monitoring Stations

• Exposure Pathway Sample Location Distance from Direction From i and/oL Sample and Designated Code + the Plant (km) the Plant _

l. AIRBORNE (Radiolodine and Particulate)

AP/CF-11 Observation Stand 0.5 NW AP/CF-12 Monroe Bridge 1.1 SW AP/CF-13 Rowe School 4.2 SE AP/CF-14 Harriman Power 3.2 N Station AP/CF-21 Williamstown 22.2 W 2, WATERBORNE

a. Surface WR-11 Bear Swamp Lower 6.3 Downriver Reservoir WR-21 Harriman Reservoir 10.1 Upriver

b. Ground WG-11 Plant Potable On-Site Well WG-12 Sherman Spring 0.2 NW

c. Sediment SE-11 Number 4 Station 36.2 Downriver From SE-21 Harriman Reservoir 10.1 Upriver Shoreline

3. INGESTION

a. Milk TM-13 Whitingham, VT 8.4 ENE TM-12 Readsboro, VT 6.1 N TM-21 Williamstown. HA 21.0 H5W

b. Fish FH-ll Sherman Pond 1.5 At Discharge and Point Inverter- FH-21 Harriman Reservoir 10.1 Upriver brates

c. Food TF-11 Honroe Bridge 1.3 SW Products TF-13 Monroe 1.9 WNW TF-21 Williamstown 21.0 WSW TV-11 Monroe Bridge ** 1.3 SW l

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. 5279R/26.217

f Table 4-1 4 (continued)

Radiological Environmental Monitoring Stations

• Exposure Pathway Sample Location Distance From Direction from and/or Sample and Deslanated Code + the Plant (km) the Plant

4. DIRECT RADIATION f GM-1 Furlon House 0.8 SW L GM-2 Observation Stand 0.5 NW GM-3 Rowe School 4.2 SE r- GM-4 Harriman Station 3.2 N _,

L GM-5 Monroe Bridge 1.1 SH 1 GM-6 Readsboro Road Barrier 1.3 N GM-7 Whitingham Line 3.5 NE Monroe Hill Barrier 1.8

( GM-8 GM-9 Duntar Brook 3.2 S

SW ,

G"-10 Cents Road 3.5 E GM-11 Adams High Line 2.1 WNW GM-12 Readsboro, VT 5.5 NNW ,

GM-13 Rcstricted Area Fence 0.08 WSW GM-14 Restrictej Area Fence 0.11 WNW GM-15 Restricted Area Fence 0.08 NNW GM-16 Restricted Area Fence 0.13 NNE GM-17 Restricted Area Fence 0.14 ENE GM-18 Restricted Area Fence 0.14 ESE GM-19 Restricted Area Fence 0.16 SE GM-20 Restricted Area Fence 0.1' SSE GM-21 Restricted Area Fence 0.it SSW GM-22 Heartwellville 12.6 NNW GM-23 W111temstown Substattor 22.2 W GM-24 b rriman Dam 7.3 N GM-25 Whitingham 7.7 NNE GM-26 Sadoga Road 7.6 NE GM-27 Number 9 Road 7.6 ENE GM-28 Number 9 Road 6.0 E GM-29 Route BA 8.2 ESE GM-30 Route 8A 9.4 SE GM-31 Legate Hill Road 7.6 SSE GM-32 Rowe Road 7.9 5 GM-33 Zoar Road 6.9 SSW GM-34 Fife Brook Road 6.4 SW GM-35 Whitcomb Summit 8.6 W5W GM-36 Tilda Road 6.6 W GM-37 Turne' rlill L ad 6.7 WNW GM-38 West Hill Road 6.6 NW C GM-39 Route 100 6.8 NNW GM-40 Readsboro Road 0.5 W

(

• Sample locations are shown on Figures 4-1 through 4-7.

" TV-11 Station is for I tafy vegetables,

+ Station IX's are ?9dicntor stations and Station 2X's are control stations (excluding the Otrect Radiation stations).

{

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( _

4-10 _ _ _ _ ________ _ _

E 5.0 SETPOINTDETgHINATIONS Chapter 5 contains the basis for plant procedures that will meet the setpoint requirements of the Effluent Monitoring Instrumentation Technical b Specifications (Spe:ification 3.3.3.6 for liquids and Specification 3.3.3.7 for gases).

(

s m

N b aevi,so, e - 2,is,ss x,,,oved 1 8 . - 4'fXm 5-1 5004R/12.243 9

5.1 Liquid Effluent Instrumentation Setpoints b

Technical Specification 3.3.3.6 requires that the radioactive liquid I effluent instrumentation in Ta' ele 3.3-8 of the Technical Specifications have alarm / trip setpoints in order to ensure Specification 3.11.1.1 is not p exceeded. That Specification limits the activity concentration in liquid effluents to the appropriate HPCs in 10CFR20, and a total noble gas HPC.

l Use the method below to determine the setpoints for the required instrumentation.

5.1.1 Method The instrument response (in counts per minute) for the limiting concentration at the point of discharge is the setpoint, denoted R and is determined as follows:

f R-( 3 (Eq. 5-1) f) (MPCg ) (Sg) where;

, f1 Flew rate past Test Tank nonitor (ggn) f2 . Flow rate past steam generator blowdoan monitor (gpm) f3 - Flow rate at point of discharge (gem)

St . Instrument response factor (cpm /(pCi/ml)) l HPCc Composite HPC for the mix of radionuclides (pCi/ml)

HPC -

ICg/JC/MPCg.Ifg/If/MPCg g g (Eq. 5-2) c i 1 i i MPCg . HPC for radionuclide i from 10CFR20, Appendix B, Table 2 Column 2 (pCi/ml)

C5 . Concentration of radionutilde 1 in mixture (pCi/ml) l

, fg . Fraction of radionuclide i in mixture

/

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Other setpoint methodologies can also be applied which are more restrictive than the approach used here.

The setpoint, R, may be set lower to accommodate pathways without on-line monitors (secondary coolant or condensste leakage). When MPC g is not stable or when dilution flow is low, R may have to be evaluated for each release.

5.1.2 Liauld Effluent Setpoint Example The effluent monitors for the Test Tank and steam generator blowdown release pathways are gamma sensitive monitors. They both have a typical sensitivity, S. of 7.5E+7 cpm per 1 pC1/ml of gamma emitters which emit one photon per disintegrasion, and a typical background of 10,000 counts per minute. Both monitors have adjustable alarm /setpoints. However, the setpo;nt adjust control is located inside the panel-mounted electronics cabinet and is not easily accessible.

[ The principal gamma emitting radionuclide in waste effluent streams is Xenon-133, averaging two orders of magnitude higher than any other specie.

However, it is not the intent of effluent monitors to respond to dissolved noble gases, because Xenon-133 concentrations have never approached the MPC.

However, Iodine-131, Cestum-134 and Ceslum-137 are detected in every liquid effluent release in roughly equal quantitles and are the principal gamma emitters because they can approach their MPCs. Therefore, for purposes of adjusting the alarm /setpoints of the effluent monitors to comply with

(

3.11.1.1, the composite MPC, MPC g

, of 6E-7 pC1/ml will be used.

It is calculated based en the following data (to be conservative Iodine is weighted greater than the Cesiums):

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1 fg MPC g Cs-134 .25 9x10-6 Cs-137 .25 2x10-5

( I-131 .5 3x10-7

[ '

MPC g -

T'i f /MPC

( q. 5-2) g

[ '

1/ (.25/9x10-6 .25/2x10-5+.5/3x10~7)

+

= 6x10'I The maximum 11guld effluent flow rate, f 3 + f 2, is taken as 130 gpm, based l on a maximum 30 gpm flow rate from the Test Tank effluent pathway and a mar.imum 100 gpm flow rate from the steam generator blowdown pathway. Both l l pathways will ba a* "iaed to operate continuously and simultaneously.

C 1 Dilution water flow, f , is taken as 140,000 gpm based on 138,000 gpm j 3

through the condenser and 2,000 gpm through the auxiliary cooling loop.

{

Throttling of cooling water is not practiced.

In this example, the setpoint for both monitors when both effluent pathways are operating is:

3 (Eq. 5-1)

{ R - (f f) (MPCg ) (Sg) 14 000 m

( = (39 g g m) (6 x 10'7 pC)/ml) (7.5 x 10*7 cpm /pC)/ml)

. 48,500 cpm

[

Note that both effluent monitors have their lower level discriminators

{ set to reject the pulses originating from the 80 kev gamma emissions from Xenon-133, and their high count rate alarms set at 48,500 cpm above background. l Revision 6 - 2/18/88 Approved By: / M M # --

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3 5.1.3 Basis The liquid effluent monitor setpoint must ensure that Specification 3.11.1.1 is not exceeded for the appropriate in-plant pathvays. The monitor is placed upttream of the major source of dilution flow and responds to the concentration of radioactivity as follows: J R Sg( fs)C gg ggy (Eq. 5-5) cpm

{

where variaules are the same as those in Section 5.1.1 except:

[ CHON Total concentration (pCl/ml) seen by the monitor sg Ratio of response from equal activities of radionuclide i to a reference radionuclide l Calibration of the radiation monitors have established that the gross gamma detector response, S f gg s was fairly independent of gamma energy, as espected. Thus, the esponse is a function of radioactivity Concentration and the gamma yield of the mixture. Since fs gg is approximately one:

R (Sg ) (C ggy ) (Eq. 5-6)

For simplicity, assume that both monitors look at the total flow for both, fg+f. 2 He know that:

f C (I+f2) f (Cggy) (Eq. 54) 3 where:

C . Total concentration at point of discharge Solve Equation 5-5 for CMON and substitute into Equation 5-4 to get:

( f 3

(Eq. 5-8)

R - (f1 f 2) (C) (Sg )

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Wa defined C = above and define MPC g such that:

{C g

&c.  ; gC i

( n. 5-,>

The right side of the equation is the MPC limit in 10CFR20, solving for

( HPC , the composite HPC for the mixture, we get the definition of MPCg :

C MPC g .

{C g (Eq 5-2)

Cj f HPC, Substituting HPCg into Equation 5-6, we get the resp:nse of the monitor as MPCg is reached at the poin's of discharge, which is the setpoint:

5 I3

) (MPCg ) (Sg ) (Eq. 5-1)

R = (f .

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'tpoints 5.2 Gaseous Effluent Instrumentati' Technical Specification 3.s.1.7 requires that the radioactive gaseous effluent instrurrentation in Table 3.3-9 of the Technical Specifications have their alarm setpoints set to ensure Specification 3.ll 2.la is not exceeded.

That Specification limits the activity concentration in off-site gaseous

( effluents to well below the appropriate HPCs in 10CFR20 by limiting total body, skin and organ dose rate, b

Use the trethod below to determine the setpoint fcr the noble gas activity monitor.

5.2.1 Nethod The noble gas activity monitor response (in counts per minute) at the

[ limiting noble gas dose (either total body or skin off-site) is the setpoint, denoted R, and is determined as follows:

R is the lesser of:

(5 ) (

ffs ) g (500) (60) l 9

R (Eq. 5-3) tb " (F) (7.83) ( t.

f g"DEFg )

and (5 ) ( (3000) (60) g=

9 { ffs ) g (Eq. 5-4)

{ R (F) (

f 0Fj) where:

(

sg Ratio of response from equal activities of radionuclide i to a reference radionuclide, i.e., Xe-133

{

0Fj = Skin dose factor (see Table 1.1-2)

- Total body dose factor (see Table 1.1-2)

( OFBI f = Fraction activity of radionuclide i to total noble gas activity

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f - - - -

( F . Primary vent stack flow rate (cc/ min)

Sg . Instrument calibration factor (cpm /(pC1/cc))

{

y Other setpoint methodologies can also be applied which are more restrictive than the opproach used here.

L 5.2.2 Caseous Effluent Setpoint Example f The primary vent stack noble gas activity monitor is an off-line system consisting of a beta sei.sitlie scintillation detector, electronics, an analog F ratemeter readout, and a digit.1 scaler which counts the detector output pulses. A strip chart recorder provides a permanent r? cord of the ratereter f output. Calibraticn data is provided by the tranufacturer which indicates the response, sg . of the beta sensitive detector to various ga',eous radionuclides. The calibration data mar verified on installation and periodically thereafter. System characteristics are:

Typical sensitivity - I cp, = 3 x 10-8 pCi/cc of Xenon-133; 7

that is, S = 3.3 x 10 cpm /(pC1/cc)

[ Typical background -

10 to 20 com Under normai plant stack flow, F, of 5.8x108 cc/ min (is 20,5% cfm x 38,300 cc/ft 3 ), one count on the scaler is equivalent to 17 microcuries of Xenon-133 noble gases released. Since the typical average primary vent stack concentrations of noble gases are only about IE-6 pC1/cc, direct grab :arpling and isotopic analysis is not satisfactory. The isotopic distribution of noble

(

gases dissolved in primary coolant is cetermined monthly and used as the distribut)on,ff,forgaseouseffluentreleases. The distribution,

{ ff,andtherelativeresponse,s,foreachradionuclideinthis g

example are presented in Table 5.2-1.

Applying Equations 5-3 and 5-4:

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b g , f3.3 x 10*7 (0.71) (500( (60 - . 73,700 cpm (5.8 x 10' ) (7.83) (2.1 x 10~ )

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(3.3 x 10' ) (0.71) (3000) (60) = 169,000 cpm

,5 (5.8 x 108) (4.3 x 10-2)

The setpoint, R, is the lesser of R tb and R sk and is, therefore,

(

73,700 cpm. This is because for the noble gas mixture in this example, the l total body dose rate is more restrictive.

[

5.2.3 Basis

{

The noble gas activity monitor setpoint must ensure that Spvcification 3.11.2.la is not exceeded. Sections 3.4 and 3.5 show that Equations 3-3 and 3-4 are acceptable methods for complying with that Sptctfication. Which b equation (i.e., dose - total body or skin) is zre limiting depenus on the noble gas mixture. Therefore, each equation must be considered separately.

The derivat't,' of Equations 5-3 and 5-4 starts with the general equation for

(

the respon;e, R, of a radiation monitor (in cpm):

R = (5 ) (

9 }ffs) g (C) (Eq. 5-5)

( (epm) (cpm /(pC1/cc)) (1) (pC1/cc) l l

or, expandir.g for the concentration:

{

R . (5 ) (

9 ffs) g (h)(60/F) (Eq. 5-10)

(cpm) (cpm /(pC1/cc))(1)(pC1/sec)(sec/ min)(cc/ min)

[ The response of the mnitor at the release rate which causes the total body dose rate limit to be reduced, Rtb, begins with Equation 3-3:

htb " 7'83 f1 0FB g substitutingh.ffhg gives: (Eq. 5-11)

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(Eq. 5-12) btb = 7.83 ff 0FB, rearranging to solve for Q:

D tb (Eq. 5-13)

{ h=

7.83Iff0F6 i

g i

Substituting Eq. 5-13 into Eq. 5-10 and substituting the total body dose rate limit gives:

S(cpm /(pCi/cc))( s)g 500 (mrem /yr) 60(sec/ min)

[ R

{ff0 Mq. 5-3) 3 3 tb

• F(cc/ min) 7.83 (pCi-sec/pCi-m ) ffG0FBg (mrem-m /pCi-yr)

The response of the nor.itor at the release rate which causes the skin l dose rate limit to be reduced, Rg , begins with Equation 3-4:

f 6,x=  ; h1 0r ,

subssitutingh=ff0hgives: (Eq. 5-11) g bgg =h ff00Fj (Eq. 5-14)

[ .

Rearranging to solve for Q:

[ . i

$' uq. 5-i5>

6= a

[ {f 0Fj

[ Substituting Eq. 5-15 into Eq. 5-10 and substituting the skin dose rate limit of 3,000 mrem /yr gives:

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F s g) 3000(mrert./yr) 60(sac / min)  !

3 (cpm /(pC1/cc))(

9 { ffC F(cc/ min) ff00Fj(mrem-sec/pCl-yr)

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L Table 5.2-1

[ Sample Calculation of Gaseous Instrumentatto.. Setpoint (Based on 1981 Yankee Data)

Detector Response Weighted Weighed Fraction With Weighted Whole Body Skin Oose

[ Noble Gas of Total Respect to Response Dose Factor Factor Specie ff0 Xe-133 1.0 s g ff0 xs g ff0 x DFB g ff0 x0F{

[

Ar-41 0.008 1.2 0.01 7.1E-5 1.2E-3 l

[ Kr-85 0.000 1.15 0 -0 -0 Kr-85m 0.010 0 0 1.2E-5 4.6E-4

{ 2.9E-3 Kr-87 0.010 1.5 0.015 5.9E-5 .

Kr-88 0.016 1.15 0.018 2.3E-4 3.0E-3

[

Xe-131m 0.020 0 0 1.8E-6 2.5E-4 Xe-133 0.38 1.0 0.38 1.1E-4 4.0E-3 Xe-133:n 0.000 0 0 -0 -0 Xe-135 0.20 1.3 0.26 3.6E-4 1.2!-2 Xe-135m 0.34 0 0 1.1E-3 1.6E-2

(

Xe-138 0.02 1.5 0.03 1.8E-4 3.6E-3

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Sumation 1.0 0.71 2.1E-1 4.3E-2

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E Table 5.2-2 )

Relatlye Fractions of Core Inventory i

[ Noble Ga$e5 After Shutdown

[ Tirre Kr-85m Kr-85 Kr-88 Xe-131m Xe-133m Xe-133 Xo-135m Xe-135 ,

t < 24 H .007 .003 .004 .004 .021 .714 .017 .2 32

[ 24 hr 1 t < 48 h -- .004 -- .005 .023 .911 .001 .056

-- .005 -- .008 .015 .971 -- --

{ 48 h I t < 5 d 5 d I t < 10 d -- .010 -- .013 .006 .970 -- --

l

.956  !

[ 10 d 1 t < 15 d -- .020 -- .022 .002 -- --

- i 151 t < 20 d -- .037 -- .034 .001 .929 -- --

l l i

[ 20 i t < 30 d -- .119 -- .071 -- .806 -- --

i

! l

-- .795 -- .103 -- .103 -- --

{30it<60d .024 .002 t1 60 d -- .974 -- -- -- --

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this header. The water from this drain header discharges without further dilution into a tributary of the Deerfleid River outside the controlled area.

[ A composite sampler collects a sample of the water whenever there is discharge (water in the pipe),

{

i Batch effluent tanks called "test tanks" collect the distillate from the 11guld radioactive waste evaporator. Normally, liquid waste accumulates at about I gpm and is processed at about 4 gpm. When a 7000 galle.1 test tank

( 1s filled, it is sampled, analyzed, and released at a nominal 30 gpm.

The condenser cooling flow provides the major source of dilution and is

[

assumed to be 138,000 gpm with two pumps operating and 69,000 gpm with one pump operating. Throttling of condenser cooling water is not practiceJ at

{ Yankee Plant.

l During shutdown periods, the 4,000 gem service water provides dilution water flow. Flow rate is variable and estimated by pump curves. Tvpi ally

( flow rates range from 1.500 to 3,500 gem. '

l The discharge rate from the $ team generator blowdown tank is fixed by

(

piping geometry and a relatively constant head on the tank. A flow meter estimates the discharge rate during periods of discharge. Verification is

{ done periodically by measuring the time for the tank level to decrease during a normal release.

The discharge rate for the lurbine Building pathway is estimated to be b 400 gpm. ,.,p .stmately I gpm of this is secondary coelant (from pump leakage and sample stations) the remainder 15 service water from various secondary plant heat exchangers. All piping is buried and inaccessible so flow is

[

estimated from cooling water pump flows, b The discharge rate for the test tanks 15 controlled by the discharge line vart ortf tces and limited to 30 gpm.

Calibration of the radiation monitors have established that the gross

( gama htector response was fairly independent of the gama energy, as expected. Thus, the response is a function of the radioactivity f .e

( Revision 6 - 2/18/A8 Approved By: Ad MM

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REFERENCES Regulatory Guide 1.109, "Calculation of Annual Doses to Man From Routine

( A.

Releases of Ret.ctor Effluents for the Purpose of Evaluating Compliance with 10CFR50 Appendix 1", U. S. Nuclear Regulatory Commission, Revision 1, October 1977.

[

8. Hamawl, J. N., "AFOLUS - A Computer Code for Determining Hourly and Long-Term Atmospheric Dispersion of Power Plant Effluents and for Computing Statistical Distributtons of Dose Intensity from Accidental Releases," Yankee Atomic Electric Company. Technical Report, YAEC-1120.

January 1977.

( C. Regulatory Guitte 1.111. "Methods for Estimating Atmospherte Transport and Dispersion of Gaseous Effluents in Routine Releases From Light-Water Cooled Reactors," U. S. Nuclear Regulatory Commission, m :* 1976.

{

D. NEo 1 and 2 Preliminary Safety Analysis Report, New Enr c t Power Company, Docket Nos. STN 50-568 and STN 50-569.

E. Yankee Atomic Technical Specifications.

F. Yankee Atomic Electric Company Supplemental Information for the Purposes

[ of Evaluation of 10CFR50. Appendix I, Amendment 2, October 1976.

(Transmitted by J. L. French - YAEC to USNRC in letters dated June 2, 1976, August 31, 1976, and October 8, 1975.)

{

G. National Bureau of Standards, "Maximum Permissible Body Surdens and Maximum Permissible Concentrations of Radionuclides in Air and in Water

( for Occupational Exposure," Handbook 69. June 5, 1959.

H. Slade, D. H., "Heteorology and Atomic Energy - 1968," USAEC, July 1968.

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APPENDIX H Radioactive Liquid Gaseous, and Solid Waste Treatment Systems Requirement: Technical Specification 6.16.1 requires that licensee initiated

{ major changes to the radioactive waste systems (liquid, gaseous, and solid) be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the Plant Operation Review Committee.

Response: There were no licensee initiated major changes to the h radioactive waste systems (liquid, gaseous, and solid) during this reporting period.

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' APPENDIX T, F

L Supplemental Information First and Second Quarters, 1988

[

1. Technical Specificatica Limits - Dose and Dose Rate Technical Specification and Category Limit

[ a. Noble Gases 3.11.2.1 Total body dose rate 500 mrem /yr 1.11.2.1 Skin dose rate 3000 mrem /yr

( 3.11.2.2 Gamma air dose 5 mrad in a quarter 3.11.2.2 Gamma air dose 10 mrad in a year 3.11.2.2 Beta air dose 10 mrad in a quarter

{

3.11.2.2 Beta air dose 20 mrad in a year

[ b. Iodine-131. Tritium and Radionuclides in Particulate Form With Half-Lives Greater than 8 days

( 3.11.2.1 Organ dose rate 1500 mrem /yr 3.11.2.3 Organ dose 7.5 mrem in a quarter 3.11.2.3 organ dose 15 mrem in a year

{

c. Lioulds 1

3.11.1.2 Total body dose 1.5 mrem in a quarter 1 3.11.1.2 Total body dose 3 mrem in a year 3.11.1.2 Organ dose 5 mrem in a quarter

( 3.11.1.2 Organ dose 10 mreta in a year l

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t APPENDIX I (Continued)

2. Technical Specification Limits - Concentration Technical Specification and Category Limit

a. Noble Gases No MFC limits

b. Iodine-131. Tritium and Radioneclides No MPC limits in Particulate Form With Half-Lives Creater than 8 days

( c. Liquids 3.11.1.1 Total sum of the fraction of MFC

{

(t0CFR20. Appendix B, Tables II, Column 2), excluding noble gases less than: 1.0 3.11.1.1 Total noble gas concentration 2E-04 uCi/cc

3. Measurements and Approximations of Total Radioactivity

a. Noble Gases "Continuous discharges are determined by indirect measurement.

Primary gas samples are taken periodically and analyzed. It is assumed that in primary to secondary leakage all gases are ejected through the air ejector. In primary coolant charging pump leakage all gases are ejected to the primary ven'. stack either during flashing or liquid waste processing. "Batch discharges" are

( determined by direct measurement. Errors associated with these measurements 6re estimated to be 155 perceut.

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L w APPENDIX I (Continued)

b. Iodines

[ Iodines are continuously monitored by drawing a sample from the primary vent stack through a particulate filter and'cnarcoal

{ cartridge. The filter and charcoal cartridge are removed and analyzed weekly. The errors associated with these c asurements are estimated to be 125 percent.

b c. Particulates

( The prticulate filter described in (b) above is analyzed weekly.

The errors associated with the determination of particulate effluents are estimated to be !30 percent.

{

d. Liquid Effluents Liquid ef fluents are determined by direct measurertent. In line

[ composite samples are analyzed for strontium - 89, strontium 90, iron - 55, gross alpha activity and carbon - 14. There is no compositing of samples for tritium or dissolved fission gas

(

analysis. For continuous discharges composite samples are used for gamma isottspic analysis. A gamrna isotopic analysis is performed on a

{ representative sample for each batch release using the Marinelli Beaker geometry. The errors associated with these measurements are

[ as follows: fission and activation products, 120 percent; tritium,

)

110 percent; dissolved fission gases, !20 percent; alpha activity,

[ 235 percent.

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r APPENDIX I (Continued)

4. Batch Releases

[

a. Liquids

[ First Quarter

[ Number of batch releaces: 11 Total time period for batch releases 4321 minutes Maximum time period for a batch release: 1815 minutes Average time period for batch releases: 393 minutes

( Minimum time period for a batch release: 165 minutes Average stream flow during period (Sherman Dam): 748 cfs Average discharge rate: 15.0 gpm

{

Second Quarter Number of batch releases: 18 Total time period for batch releases: 7586 minutes Maximum time period for a batch release: 1824 minutes

( Average time period for batch releases: 237 minutes Minimum time period for a batch release: 421 minutes Average stream flow during period (Sherman Dam): 442 cfs

{

Average discharge rate: 18.0 gpm

b. Cases There were no batch releases during the first and second quarters.

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4824R/20.124 )

i

u APPENDIX I (Continued)

5. Abnormal Releases

[

a. Liquid

[ There were no nonroutine liquid releases during the reporting period.

b. Gases

[ Ihere were no nonroutine gaseous releases during the reporting period.

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Tstsphons (617) 872-81C0

• TWX 710-380-7019 YANKEE ATOMIC ELECTRIC COMPANY 1671 Worcester Road, Framingham, Massachusetts 01701 ys y August 30, 1988 FYR 88-117 United States Nuclear Regulatory Commission I Document Control Desk Washington, DC 20555 References (a) License No. DPR-3 (Docket No. 50-29)

Subject:

Semiannual Effluent Release Report

Dear Sir:

Enclosed are the tables summarizing the quantities of radioactive liquid and gaseous effluents, and solid waste released from Yankee Nuclear Power Station at Rowe, Massachusetts for the first and second quarters of 1988. This information is submitted in accordance with Technical Specification 6.9.5.b.

We trust that this information is satisfactory; however, should you have any questions, please contact us.

Very truly yours.

YANKEE ATOMIC ELECTRIC COMPANY o^

p 0 orge Papanic, Jr.

Senior Project Engineer - Licensing GP/25.792 cc: USNRC Region I USNRC Resident inspector, YNPS i t i

l