ML20063M155

Rev 0 to Fsv Nuclear Station Decommissioning Project Final Survey Plan for Site Release

ML20063M155
Person / Time
Site:	Fort Saint Vrain
Issue date:	02/17/1994
From:	SCIENTIFIC ECOLOGY GROUP, INC.
To:
Shared Package
ML20063M152	List: ML20063M150 ML20063M155
References
PROC-940217, NUDOCS 9403100069
Download: ML20063M155 (72)

Text

m 4

MQmfRL:?,ga.::

=- GR7dia'? aws,u gg;,si,,,*'"MLaw;,,;;mu,L

nszwgaww = i-

,WJ Cg,gwg2; g Fort St. Vrain Nuclear Station Decommissionmg Proj. ect Final Survey Plan for Site Release

.....z g====armw=mm l-;,

t~rmmntm wwnwahw WWmmy!?mn aaecduaxsad p

Revision 0

?j 73

. g"y W WQ W Z> fB 97m wsm gj j premmegg f

dj Sj 4,m asisdhd February 17, 1994

))

q d

'J

,2f? " '_ A j j u fj j g 7 q gm+ m;.y?

(gif an",

4 O

t

+

s

s,

?

^'

ll

ed} A$_.. l3sammazd Prepared by:

1 I

I I

I Scientific Ecology Group /RE&DS I

I Radiation Protection Department l

l Fort St. Vrain Nuclear Station j

SCIENTIFIC ECOLOGY GROUP, INC.

6============n--~

_==u==-===---~====w=n-~

- mi 9403100069 940217 PDR ADOCK 05000267 W

PDR j

~

Fort St. Vrain Nuclear Station

[

Decommissioning Project Final Survey Plan for Site Release Revision 0

[

[

[

[

Prepared by:

[

Scientific Ecology Group /RE&DS

[

Radiation Protection Department Fort St. Vrain Nuclear Station i

[

[

[

[

[.

]

s FSV FINAL SURVEY PLAN PREFACE PREFACE This document describes the methods used by the Public Service Company of Colorado (PSC) n demonstrate that radiation and radioactive contamination levels at the Fort St. Vrain Nuclear Station (FSV) have been reduced to levels below criteria established for unrestricted use.

This plan has been developed in accordance with the Fort St. Vrain Decommissioning Plan (DP), FSV Decommissioning Technical Specifications (DTS), Draft NUREG/CR-5849,

" Manual for Conducting Radiologitai Surveys in Support of License Termination," and 10 CFR 50. It supplements and updates the description of the proposed final radiation survey presented in the Fort St. Vrain Decommissioning Plan.

[

This plan has been developed as administrative guidance.

It is intended to provide the basis for the implementing procedures governing the conduct of the final survey.

Revisions to this plan will be controlled in accordance with 10 CFR 50.59.

This plan includes a description of the technical considerations and methods to be used for design and implementation of the final survey. The methods described are derived from regulatory guidance, specifically Regulatory Guide 1.86,

" Termination of Operating Licenses for Nuclear Reactors," Draft NUREG/CR-5849, " Manual for Conducting Radiological Surveys in Support of License Termination," NUREG/CR-5512, " Residual Radioactive Contamination From Decommissioning," and from recent U.S. reactor facility decommissioning experience (U.S. Army Materials Technology Laboratory, Shoreham Nuclear Power Station, Pathfinder, Saxton, Shippingport, and UC Berkeley), taking into

{

account conditions that are specific to the Fort St. Vrain facility.

Note: For the definitions of terminology used in the Final Survey Plan, please refer to Section 8.0, " Glossary."

1 1

r 2/17/94 REVISION 0 r

t FSV FINAL SURVEY PLAN TABLE OF CONTENTS I

TABLE OF CONTENTS SECTION TITLE PAGE 1.0......... HISTORICAL BACKGROUND...................

1-1 2.0......... SITE INFO RM ATION......................... 2-1 2.1......... Site Description............................. 2-1 2.2......... Site Conditions for Final Survey................... 2-2 2.3......... Scope of Final Survey......................... 2-3 3.0......... FINAL SURVEV OVERVIEW................... 3-1 3.1......... Survey O bjectives............................ 3-1

[

3.2......... Identity of Contaminants....................... 3-4 3.3......... Determination of Site-Specific Limits................ 3-5 3.3.1.

Alpha and Beta-gamma Removable Surface Contamination

(

Limits.............

.3-6 3.3.2 Alpha and Beta-gamma Total Surface Contamination Limits.....

. 3-6 3.3.3.

.. Gamma Exposure Rate Limit (External Effective Dose) 3-6 3.3.4

.. Hard To Detect Nuclide Limits

. 3-7 3.3.5 Soil and Water Activity Limits........

.... 3-7 3.3.6

. Admiristrative Action Levels..........

.. 3-7 3.4......... Organization and Responsibilities.................. 3-8 I

3.5......... Quality Assurance............................ 3-9 3.5.1........ General Provisions....................

....... 3-9 3.5.2 Final Survey Quality Control Procedure.....

......... 3-11 i

[

3. 6......... Trai nin g.................................. 3-12 l

3.7......... Laboratory Services........................... 3-12 3.8......... General Survey Plan.......................... 3-12

[

3.8.1

.. History File Preparation...

................. 3-13 3.8.2.

.... Initia1 Classification

........ 3-13 3.8. 3...

.. Walkdown.......

... 3-14 I

(

3.8.4.

. Tumover for Final Survey

.............. 3-14

3. 8.5...

Survey Design.

.................. 3-15 3.8.6...

.... Modifications............................... 3-15 l

[

3.8.7.

.. Work Planning and Scheduling....................

.3-16 3.8.8..

... Survey Instmetions.................

.......... 3-16 3.8.9..

... Field Support

.3-16

(

3.8.10.

... Survey Measurements..........

............ 3-16 3.8.11

.... Restoration and Isolation

....................... 3-17

3. 9......... S ch e d ul e.................................. 3-17 3.10........ Final Survey Reporting

........................3-18 2/17/94 i

REVISION 0

FSV FINAL SURVEY PLAN TABIE OF CONTENTS SECTION TITLE PAGE 4.0......... SURVEY PLAN AND PROCEDURES.............. 4-1 4.1......... General.................................. 4-1 b

4.2......... Instrumentation............................. 4-1 4.2.1.

Instrument Detector Description..............

.. 4-2 4.2.2

....... Detection Sensitivity......

............. 4-2 4.2.3 Minimum Detectable Activity Calculation..........

4-5 4.2.4 Calibration and Maintenance,......

...... 4-5 4.3......... Su rvey Plan................................ 4-5 4.3.1........ Classification

.... 4-5 i

4.3.2.

Reference Locations

. 4-11 1

I 4.3.3

... Survey Maps.........

..... 4-11 4.3.4

... Survey Point Identincation

. 4-12 4.3.5..

..... Surface Scans.........

.. 4-12 4.3.6 Surface Activity Measurements

.... 4-12 4.3.7...

... Exposure Rate Measurements

..... 4-13 4.3.8 Soil and Water Sampling

..... 4-13 4.3.9...

... Special Sampling and Measurements

..... 4-13 4.3.10..

.... Sampling for Hard To Detect Nuclides.

...... 4 14 4.4......... Background Level Determination..................4-14

(

4.4.1

... General Requirements....

.4-14 4.4.2..

..... Objectives of Background Determination

. 4-14 4.4.3...

.... Background Measurements.............

......... 4-15 4.4.4

.. Documentation And Control of Background Measurements 4-17 4.5......... Sample Analysis............................. 4 18 5.0......... DATA INTERPRETATION

.....................5-1 5.1......... Conversion of Measurements to Reporting Units........

5-1 L

5.1.1.

. Direct Measurements - Total Surface Activity

. 5-1 5.1.2..

. Removable Contamination Measurements 5-1

{

5.1.3..

.. Gamma Exposure Rate Measurements

. 5-1 5.1.4....

. Soil and Water Activity Measurements...

5 -2 5.1.5

.. Ilard to Detect Nuclide Measurements....

5 -2 I

5.2......... Comparison With Release Limits.................. 5-2 5.2.1

. Attainment of Release Limits for Surface Contamination.....

5-2 5.2.2

.... Attainment of Gamma Exposure Rate Limit...

5-3 r

L 5.2.3 Attainment of Soil and Water Limits................. 5-4 5.2.4.

. Attainment of Ilard To Detect Nuclide Limits...........

5 -4 6.0......... FINAL SURVEY REPORTING................... 6-1 6.1......... To pi cal O u t li n e..............................

6-1

?

2/17/94 ii REVISION 0

I FSV FINAL SURVEY PLAN TABLE OF CONTENTS l

SECTION TITLE PAGE l

6.2......... Reporting of Survey Findings

....................6-3 6.2.1...

... Summary...........

............. 6-3 6.2.2....

Summary Data Reporting for Each Survey Unit 6-3 6.2.3...

Detailed Data Reporting

.6-4 7.0......... R eferen ces................................. 7-1 j

l 8.0......... G lossary..................................

8-1 r

1 l

TABLE LIST OF TABLES PAGE l

l Surface Contamination Limits

.. 3-3 3.1 l

4.1 Final Survey Instrumentation Summary

. 4-3 4.2

. Typical Detection Sensitivities for Tc-99 and Th-230.......

4-4 i

4.3..

Initial Facility Survey Area Classification....

. 4-7 i

FIGURE LIST OF FIGURES PAGE 2.1 Facility Layout Map

......... 2-4 4.1......... Survey Unit Classification Process

. 4-19 4.2.

.... Reference Grid Layout..

......... 4-20 4.3

.... Final Survey Unit Map (Example).

...... 4-21 l

5.1

.... Total Surface Contamination Data Processing Affected Survey Units....

............ 5-5 j

5.2.

. Total Surface Contamination Data Processing Unaffected Survey Units

...... 5-6 1

5. 3.....

Removable Surface Contamination Data Processing........

5-7

.i a

5.4

. Gamma Exposure Rate Data Processing 5-8 5.5...

Soil Data Processing.....

..... 5-9 t

I 2/17/94 iii REVISION 0 u

~

FSV Final Survey Plan HISTORICAL BACKGROUND 1.0 HISTORICAL BACKGROUND I

Public Service Company of Colorado (PSC) operated a 330 Mw(e) High Temperature Gas Cooled Reactor (HTGR) from July,1979 until August,1989. The plant, designated as the Fort St. Vrain Nuclear Station (FSV). is located approximately 35 miles north of Denver and three and one-half miles nonhwest of me town of Platteville in Weld County, Colorado.

Construction of FSV was authorized by the U.S. Nuclear Regulatory Commission (NRC) by issuance 'of a provisional construction permit on September 17, 1968.

Construction was l

completed in December,1973 and a facility operating license was granted on December 21, 1973. Initial fuel loading commenced on December 26,1973 and initial criticality was achieved i

January 31, 1974. After a prolonged period of startup testing, low-power operation and plant j

modifications, the plant was committed for commercial operation on July 1,1979. Full power j

was achieved November 6,1981.

l FSV was a load-following central station power plant which used a High Temperature Gas-Cooled Reactor (HTGR) to produce steam for the generation of electric power. The reactor I

design utilized the same fundamental principles that formed the basis of the 40 Mw(e) prototype HTGR at Peach Bottom, Pennsylvania.

That prototype began supplying power to the l

I Philadelphia Electric Company system in March of 1967; began commercial operation on June 1,1967; and ceased operation on October 31,1974.

I In the nuclear steam supply system for FSV, heat was produced by fission in the HTGR utilizing a uranium-thoriuru fuel cycle. Graphite was used for the moderator, core structure, and

)

reDector. High temperature helium was used as the primary coolant to produce superheated and reheated steam at a temperature of 1,000 "F to match conventional thermal station conditions.

The entire nuclear steam supply system, including the reactor core, graphite moderator and reflector, steam generators and helium circulators, was contained within a Prestressed Concrete Reactor Vessel (PCRV).

The active core was composed of 1,482 hexagonal graphite fuel elements stacked in 247 vertical I

columns. The fuel was in the form of panicles made of a mixture of the carbides of thorium and uranium which was coated with highly retentive coatings of pyrolytic carbon and silicon carbide. The fuel particles and a carbonaceous matrix form bonded fuel rods which were located I

within the fuel holes. The combination of a relatively large fuel particle (200-500 microns) and a four-layer coating provided the particles excellent fission product retention characteristics.

During the timeframe of operation, FSV operated for approximately 890 effective full-power days. FSV was shutdown on August 18, 1989. On August 29, 1989, the PSC Board of Directors reviewed and confirmed the Executive Management decision that FSV would not be restarted, and that PSC would pursue decommissioning of FSV. The decision to permanently j

shutdown and decommission FSV was based on related technical and financial considerations.

Problems were identified with the control rod drive assemblies and the steam generator steam ring headers that presented significant technical obstacles which could be overcome, but at i

signiDeant cost in dollars and time to PSC. In addition, due to the uniqueness of the one-of-a-2/17/94 1-1 REVISION 0 E

l FSV Vinal Survey Plan IIISTORICAL BACKGROUND

}

kind HTGR fuel cycle, the cost to purchase new fuel was prohibitive. This, in conjunction with low plant availability and correspondingly high operating costs, made continued operation of FSV imprudent.

A preliminary decommissioning plan based on the SAFSTOR alternative was filed with the NRC in June,1989. Subsequent evaluation by PSC resulted in the decision to select the DECON j

alternative, and the Proposed Decommissioning Plan (PDP) was filed with the NRC in November,1990. PSC's objective is the dismantlement and deconunissioning of FSV to release all site areas for unrestricted use. To accomplish this, a portion of the PCRV structure and the radioactive balance-of-plant equipment which exceed the limits for unrestricted use will be decontaminated or removed as described in the Fort St. Vrain Decommissioning Plan.

y L

in May,1991 a 10 CFR 50 Possession Only License was granted by the NRC.

The DECON alternative has been chosen for implementation of decommissioning activities at FSV. This involves the decontamination and dismantlement as necessary of plant systems and areas to allow release of the facility for unrestricted use. In the Generic EnvironmentalImpact L

Statement on Decommissioning Nuclear Facilities (GEIS) (Ref.1), the NRC concludes that I) ECON is advantageous if the site is required for other purposes or if the site is extremely

{

valuable. DECON is advantageous to FSV because of its potential conversion to a gas-fired generating station and because the FSV switchyard is valuable for power distribution within the PSC system. The current regulatory process for DECON is also well understood and the choice

{

of DECON avoids the need to plan for contingencies to accommodate new requirements which may be introduced in future regu!atory changes. Additionally, the operating staff is available to assist with surveillance and maintenance of the facility during decommissioning activities, u

m u

l I

E 2/17/94 1-2 REVISION 0

FSV FINAL WJRVEY PLAN

}lITE INFORMA TION 2.0 SITE INFORMATION 2.1 Site Description The Fort St. Vrain Nuclear Station is located approximately 35 miles north of Denver and 3.5 miles northwest of the town of Platteville in Weld County, Colorado. The site is located in an agricultural area with gently rolling hills. Grade elevation at the plant is 4,790 feet above sea level.

The 2,798 acre site owned by PSC is identified as the Owner-Controlled Area. Farming has been continued on Owner-Controlled area of the site, but there are no farming operations or permanent residences located within the Restricted Area. The Restricted

(

Area is surrounded by a Security fence, and access is controlled for purposes of l

protection of individuals from exposure to radiation in accordance with the Decommissioning Plan (DP), Section 8, Decommissioning Access Control Plan (Ref. 2).

l I

The station is located approximately two miles south of the confluence of the South Platte

)'

River and the St. Vrain Creek. Neither of these two streams are considered navigable.

(

Cooling for the plant is provided by mechanical draft cooling towers. Make-up to the cooling towers is obtained from the two streams, and is supplemented by shallow well water. Nineteen shallow wells are located on the site. The licensee also owns surface water rights in four irrigation ditches which traverse portions of the site.

1 A considerable amount of data regarding the levels of radiation and radioactive material in the environs has been collected and reported in the annual Radiological Environmental Monitoring Program (REMP) Reports (Ref. 3 & 4). The data indicates that there is little

{

probability that any significant amount of radioactive liquids could have entered the groundwater during power operations. Any effluent which was discharged from the radioactive liquid waste system was diluted by the cooling tower blowdown line prior to

[

release to surrounding surface waters. There were no spills or accidents during operation 3

of FSV which had the potential of contaminating the site. The REMP will be continued

{

throughout the decommissioning to ensure that any contamination of the site environs

{

(

which could occur as a result of decommissioning activities is detected.

j Liquid effluent is released from the FSV site via the Goosequill Ditch. From the

(

concrete-lined Goosequill Ditch, liquid effluent flows into the Jay Thomas Ditch, where additional dilution may occur, and then on to a 25 acre farm pond that contains approximately 32 million gallons of water. The total distance from site discharge to farm

(

pond discharge is approximately 8,700 feet. An alternate flowpath, via a slough, to the St. Vrain Creek and then on to the South Platte River is utilized when discharge via the Goosequill Ditch is not possible.

Within the area surrounding the site, there is no evidence of recent earth movement along any known faults. Shales of the Pierre Formation form the bedrock at the plant location.

Subsoils at the site generally consist of a thin layer of sands, capable of supporting light f

2/17/94 2-1 REVISION 0

l FSV FINAL SURVEY PLAN SITE INFORMATION l'

stmetures; underlain by medium dense sands becoming gravelly with depth that can support moderate loads; and thence by hard to very hard claystone bedrock, capable of g

[

supporting heavy loads, found at a depth of 44 to 54 feet, The major structures within the Restricted Area include the Reactor Building which r

L contains the Prestressed Concrete Reactor Vessel (PCRV), Turbine Building, Radwaste Compactor Building, New Fuel Storage Building, Technical Support Building which c-contains the Radiochemistry Laboratory, Mechanical Draft Cooling Towers, Warehouse L

and Construction Workshops, Evaporation Ponds, and the Electrical Switchyard. The ground surface covering within the Restricted Area is composed primarily of gravel and vegetation, with smaller portions devoted to concrete or asphalt roadways and laydown r

areas.

The NRC issued PSC an Environmental Assessment of No Significant Impact for the L

constmetion and operation of an Independent Spent Fuel Storage Installation (ISFSI).

The ISFSI is located on approximately 20 acres of the Owner-Controlled land outside of the Restricted Area for the plant, approximately 1,500 feet northeast of the Reactor Building. The ISFSI facility is operated under a separate license in accordance with 10 CFR 72, and is not included in the scope of this decommissioning.

2.2 Site Conditions for Final Survey The radiological status of the FSV facility and environs has been evaluated in the FSV Initial Radiological Site Characterization Report (Ref. 5) in order to provide information pertinent to the decommissioning and fmal survey of FSV. Additional radiological information has been provided by the REMP; the Environmental Radiological Surveillance Program (ERSP); and the study conducted by Colorado State University's Department of Radiology and Radiation Biology from June,1990 until October,1990 (Ref. 6) to assess the pre-operational radiological status of the ISFSI site.

The FSV facility will be largely left intact following decommissioning. Dismantlement of structures will be confmed to the PCRV (Prestressed Concrete Reactor Vessel), and portions of the Reactor Building, Turbine Building, and Liquid Waste System. Removal will be for purposes of removing contaminated structures and to provide paths for removal of contaminated piping and equipment.

Following defueling, the PCRV contained the majority of the remaining radioactive material inventory. Portions of the PCRV concrete are activated due to direct irradiation from the reactor core, and will be removed prior to Final Survey and disposed of as radioactive waste at a licensed radioactive waste disposal facility.

To date, seventeen balance-of-plant systems have been identified as being contaminated in excess of the limits for unrestricted use. All piping and equipment contaminated in excess of the limits for unrestricted use will be decontaminated and left in-place, decontaminated and free-released, or dismantled and removed from the facility and disposed of as radioactive waste at a licensed radioactive waste disposal facility.

2/17/94 2-2 REVISION 0

FSV FINAL SURVEY PLAN SITE INFORMATION The REMP will be continued to ensure that any contamination of the site environs which could occur as a result of decommissioning activities is detected.

2.3 Scope of Final Survey The Final Survey will include all pertinent structures, surfaces, systems and components, concentrating on those previously identified as contaminated or potentially contaminated and those identified as contaminated or potentially contaminated during the dismantlement / decommissioning phases. The final survey will include:

Sampling outside the restricted area of PSC property, soil, pavement, water, and liquid effluent ditch and pond sediment for radioisotopic analysis and measurement of gamma exposure rate, Sampling inside the restricted area of PSC property, soil, basin sediment, pavement and water for radioisotopic analysis and measurement of gamma exposure rate, Radiological surveys of the PCRV and Reactor Building, and I

Radiological surveys of the Turbine Building, Radwaste Compactor Building, New Fuel Storage Building, Radiochemistry Laboratory, Helium Transfer and Storage System, and Liquid Radwaste System.

I I

l I

l l

1 l

)

i

)

il l

l 2/17/94 2-3 REVISION 0

h I

$l-

.f,i _,,_,,_,3.

,1

+

...u

. al 7 _ ___ _ _

d

,,- 4 w e, 6 m w l

.e t.... -.

i j:n

,+ f.

e..

~w~-

s.,:m ~.

= = *

.q xs r

1,_'_.',,s*~ e, ->:i:

L c,,s,

~

-.7

,,, i U,-

s

.].

1 g

1. r~.-7 x 6

a

/

i 1

- ~ s t.;) r 6

i 1

r f

d-i a

(. j-a s.-

Mj,a :!!nI c, '

N q,k

~

s W

IW m.p;p;b

. t. @u s

=== = - - -

y M;p l i..p"t If'=I -

---,i i,

a i p e

a

,. a i 2 F.

a

.i r

J1.

a i

w. vy Asc :

o

. wn Mg,f;w[7A.f,p!

i i

t e

14 oi ai

..n 4pM L

.h i

b:.

4 e.

t !(l;g[

4l;l j ll 61 ti i i

t; io M

i[ [ f W n idM W

g

' "@: trrr 1Mi iL i:m

~ !l, qiy l lll h

t I

m iIn i; me ll t

olb.i._ _ _ _

._.a j j

. eei i

n l

j= s. ; g;

, ).k 5@."* i U I m _..-. - _._ _

, J'

:n i

,--------,s, o

g. _ _ _ _ _. _

_ j o

...... n: y:j i i, $u i

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

--- n e,1 )

a r.: l

_. _. _ _. -. - - - ~- - - -5 l.J yl

}

ie r.

>i

.j

(

,li 7

l,

_s,...,

i

^'

s t,

p yll mm

~~ ~ -'~- [s(s d u _ _ _.... _ _

, jl.j tt

,p...

y

[

q; j-r gg-(.

3u3 r_

y c

ll I '

~

4 l

l,

ill 3

I L

'[

l

[ [a '

I i.

i 4

[

i lUl u

3 c.

1 L

.6 3

3 i

f gl f

[._.__._-.d.-

k g

e i

1.__.m.

.L l

[

n } :i a

_ _. w.m, n

l

=

w Q

,i

)

i

(; ?, pp:.f / ~._. _.

s-e p

,I l

n _ _ _._ _ _ _ _ _ _. a

..-.- +.;

e

,, s y;;

L 1,s j

.; t t i

y.

,e u.

g _l..

,s.j/

[t f--*

F f, /,

V. ' ~

e r.

3 L.'

- ~

j, la7./.

,,;,w,,,-

4

.g 9,

j-x 4,-

) n</

W c ll 5vQ1[s U

.e l

/

g s3 '\\

,y y

w j.

x n

h - & i") k j

,1-_...,.,.

k.

.. W o

i i

ti ll 1

l 21 4

~

i i

l l

e si

.l

-I j

s

(...

(,'o l

3 I

ya gi t

iii. H j

t(

y

/

'.,./ /

y n

/ /

Q

+

^

j }g h {-

.- x

.m w.. _ _.__._ _ _.. _ a n w o s .. _.___.-_~~._._-.-..7-------1,- o ,.... ~ - -... ' ' * * ' ~ ~ " ' * ~

• ' ' ' ' - ~ ~ ~ ~ ' ~ ~

_..,s_ -.-. d.~..-~,~- ~_... _,-g.= _- _ ; 7; - = { w,,,,, c, _~ -)- s g) -~~~~' p~r --"*L__.....s.---.-----.----~--.--~-~~~~~ / W . - m... . ~.m.: ' j -:.': p'" ~ 7<, p p. - b g .. h.3 p m i r } t 1; n.. ..s

a r= = = = = = =_. h _d' = ; ;== u - b ~===.2 n = = n u;u [.=x=:=:- ~. g

1===

= wi;L An g,aI.- y,. ~W s umn u m f, i E! M. _.4.t. ii n ,e i o_ r w ,,i / m (lp,

r-

]} s -- 1 1 c f 9 ,w o. _. -. w. i n" s 0? l Q ll \\ e-.~. I:' / M s\\ li \\ a m;" ll Ufl 7 ? I "s..y /I, i; .: a af L s ];*! I 3 b .N ,il h,1.,s.7 i,- ,,z.., a, \\ .~s N, N t l _(u, g8f Il J..W1. p* i ~ v.. gg 4-- - ANSI.EC; xN u.x 3 [. ] 2 ii m - "p,E_ i c lh APERTUREg h:P ttr : W h a4, l l CARD l a-p,r,y i, J M. p w g-t__p.yl$,i i O jI D. m.e um. 'c.. !! Also Available on .J. d k Aperture Card Yn H } <', ~ l.3. : rr _.. k_ tr. / 1+ri ' f%l g{'-,, -t--c:.c--! (7{f.II.qlih4 h ; p!' I l ~ l il.1p,y r~ g on s _ _..' I,; n l y

I '7.s, A.

lj c

sl' b

y, I . fd). a 4 l'. p'!. I] j] /!mm,.a.f;nQ, d BRB BHB BDH Mt f.. s if / r 4(>s 33 D f... egf.' jgf, e a. m '.+%,y'4}}a'?s:[$ hse hl ]$@ @]}B @N@ .] s i j ) 1. n,t..... + . p w' w: e. ,i h r { ,b h[ o n\\ ,'N. n n}e, e n g.. t.i ~ gag a,nB .,_-- n i.i ~ (, l .a w c g:rB q m. ~ .I y a... 41., m arc au sa:e p e.~ s i _a ,mu wm.g~ii;y (? p_A.g. e n e:x _.b. j qqg p m g. r,s r -p=,~x Ti~~ ~~~ <rr d l J1K, y p+y 3..g;p.yt v.- . m d_, w w ~' s I; (~ $ t r. - 1 l l ~.. .g

i

}/ 3 s Q; / n-... y,p y g g>y @j ja >{((-w {* [ e'~ ] I b i R j' ~ ~ f" fi w i I

t _

.

i c . (<: e. ) \\ I i[ 'n\\ \\ II h _h /,/>, ' \\m i y,l g, s r. ' le l l Ns i I 2 l 'I l..,g 2_=...--. q:e // e jp/,, ~q, r:,, s m,.. m e l{ g) f=&- ,/ 7, / ..d,! ~

t. )I-

.. 5 ,1 y 3 'fr i

l. p

[u,iW! [ , /,/ nl a _x j ) y _, g,,,,, -. - {. q -_, q p. _, __.a(7 .--.(.- g, .o, ( ti; L T. a.. e. ;,,,,, 1 .,q ,,,,, 4 L l f ) \\ i s ,f,/w. I {.. 'y ~~1,,,_,,,, ( J ,I J. l1,, n.>.,1 jj j p t,j j .y g., _.g g gp .. s 1 ji 4 u ; . l, b= a c-1,i n g [ ((. = i ) 4 . ' '. ] r .c C i,l l ll 11 l fl m 4 - d I t i^ 7 ( % y. i, g g

ie 2

rs i il-ai e < 1( 3 n: p3 i 3 ;. I,i i os. i _ r t,: .. J g 4 p l l I ) i ; l '; WI 1 Hl l; (1 j - -..Q l} t %O3lOOO M-9Y.t 4:.m e#igst.--$. ,.w-.=g9 y i.-irF_y.w -emtyw e-qt +r'e s g -4Ft':w-EU'-N

[ FSV FINAL SURVEY PLAN FINAL SURVEY OVERVIEW k 3.0 FINAL SURVEY OVERVIEW A general overview of the Final Survey Plan and its objectives is provided in this section of the 7 plan. Specific details of the methodology to be used during the final survey are provided in l subsequent sections of this plan. 3.1 Survey Objectives [ The Final Survey is designed to demonstrate that licensed radioactive materials have been ( removed from the Fort St. Vrain (FSV) facilities and property to the extent that residual levels of radioactive contamination are below applicable limits in Regulatory Guide 1.86, " Termination of Operating Licenses for Nuclear Reactors," (Ref. 7) and that Public ( Service of Colorado (PSC) has met all necessary requirements for termination of the current 10 CFR Part 50 License. [ The surface contamination limits for beta-gamma emitting nuclides presented in Table 3.1, " Surface Contamination Limits," will be used to measure the effectiveness of decontamination and dismantlement efforts. Radiation detection instrumentation selection for perfonning direct measurements of total surface contamination during the Final Survey is based upon the conclusion that the most readily-measured radionuclides in determining compliance with Regulatory Guide 1.86 release limits are activation products dominated by Co-60 and Eu-152. These are beta-gamma emitters as defined in Regulatory Guide 1.86, hence the limits for beta-gamma emitters shown in Table 3.1 apply. The surface contamination limits for all classifications of alpha emitters are also shown in Table 3.1. Although there is no history of fuel failure and significant quantities of alpha emitters have not been detected in plant contamination deposits, the past presence of irradiated fuel in the facility provides a potential source of alpha contamination. Methods and instmments are included in the Final Survey Plan to reconfirm the appropriate alpha classification and contamination limits contained in Regulatory Guide 1.86, and to measure alpha surface activity at levels below the appropriate limits in Table ( 3.1. The gamma exposure rate, as measured at a distance of one meter from accessible l [- surfaces in facility buildings and outdoor areas, will be limited to 5 R/hr above the background. The background gamma exposure rate will be determined for each set of measurements in each survey unit. The methodology used to establish gamma I ( background is described in Section 4.4.3.f. Gamma exposure rate values exceeding 5 pR/hr above background will warrant additional measurement and evaluation for remediation as discussed in Section 5.2.2. c L Where residual contamination resulting from liard To Detect Nuclides (IITDN) is suspect, an analysis will be performed to evaluate the Total Effective Dose Equivalent r 2/17/94 3-1 REVISION 0

~ FSV FINAL SURVEY PLAN FINAL SURVEY OVERVIEW (TEDE) from IITDN. A more detailed description of this analysis is provided in Section 3.3.4. Prior to, or during early phases of the Fina! Survey, representative samples from the locations described in Section 2.3 will be acquired and analyzed. The results of these analyses will be used as necessary to reconfirm the: [ Radionuclide inventory Significance of Ilard to Detect Nuclides in residual contamination [- Appropriate acceptable alpha surface contamination levels Methodology and instrumentation required to perform final surveys [ l 1 f 1 l I L f L I l L h I ( 2/17/94 3-2 REVISION 0

' FSV FINAL SURVEV PLAN FINAL SURVEY OVERVIEW TABLE 3.1 [ SURFACE CONTAMINATION LIMITS ( 6 6d b NUCLIDE' AVERAGE' MMIMW . REMOVABLE ' 2 2 2 U-nat, U-235, U-238, and 5,000 dpm a/100 cm 15,000 dpm a/100 cm 1,000 dpm a/100 cm associated decay products 2 2 2 Transuranics, Ra-226, Ra-228, 100 dpm/100 cm 300 dpm/100 cm 20 dpm/100 cm Th-230, Th-228, Pa-231, Ac-227, I 125, I-129 2 2 8 Th-nat, Th-232, Sr-90, 1,000 dpm/100 cm 3,000 dpm/100 cm 200 dpm/100 cm Ra-223, Ra-224 U-232, [ I-126, 1-131, 1-133 2 2 2 fleta-garnma emitters (nuclides 5,000 dpm By/100 cm 15,000 dpm py/100 cm 1,000 dpm Sy/100 cm 7 with decay modes other than [ alpha emission or spontaneous lission) except Sr-90 and others noted above.

• Where surface contamination by both alpha-and beta-gamma-emitting nuclides exists, the limits established for alpha-and beta-gamma-emitting nuclides should apply independently.

'As used in this tabic, dpm (disintegrations per minute) means the rate of emission by radioactive material as determined by correcting the counts per minute observed by an appropriate detector for background, efficiency, and geometric factors associated with the instrumentation. ' Measurements of average contaminant should not be averaged over more than I square meter. For objects of less surface area, the average should be derived for each such object. "The maximum contamination level applies to an area of not more than 100 cm. 2 2 'The amount of removable radioactive material per 100 cm of surface area should be determined by wiping that area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instrument of known efficiency. When removable contamination on objects ofless surface area is determined, the pertinent levels should be reduced proportionally and the entire surface should be wiped. Note: Table 3.1 limits are based upon the acceptable surface contamination levels provided in Regulatory Guide 1.86. ] i 2/17/94 3-3 REVISION 0

VSV FINAL SURVEY PLAN FINAL SURVEY OVERVIEW [, 3.2 Identity of Contaminants ( An estimate of the gross radionuclide inventory was completed in June,1989 and was included in the Supplement to Applicant's Environmental Report (Ref. 8). The total inventory (source term) due to activation and activated corrosion products three years [' after shutdown was estimated to be 1.3 x 106 curies (Ci), not including the irradiated fuel. Of this amount, approximately 99.6% was estimated t) be contained in the activated and contaminated portions of the PCRV, which will be removed prior to conducting the final survey. The remainder of the inventory was estimated to be contained in balance-of-plant surfaces and system components and constitutes 3 approximately 0.4%, or 5.2 x 10 Ci, which will also be decontaminated or removed prior to conducting the final survey. [ The design of the HTGR results in a source term which is unique to FSV. This source term is primarily a result of neutron activation of both metallic and concrete components of the PCRV and neutron activation of impurities contained in graphite components of [ the PCRV. These activation products include both beta-gamma emitting radionuclides such as Co-60, Eu-152 and Eu-154 which are the major contributors to whole body dose, as well as HTDN such as H-3, C-14 and Fe-55. The latter, although insignificant with respect to the TEDE, are significant in terms of the overall radionuclide inventory. The Decommissioning Plan indicated that the activity in the graphite components of the l PCRV is dominated by Fe-55 and H-3, which were generated due to neutron activation of impurities in the graphite. Due to the large volume of graphite and the higher specific activity of Fe-55 and H-3, these two nuclides are the largest contributors to the overall L radionuclide inventory which will be removed prior to conducting the final survey. Although not the dominant nuclide in terms of radionuclide inventory, Co-60 is the dominant nuclide contributing to the TEDE in the graphite components. L The dominant nuclides for stainless steel components are Co-60, Fe-55, Ni-63, Ni-59 and [ Mn-54. The dominant nuclides for carbon steel components are Co-60, Fe-55, Ni-63, l and Mn-54. Overall, the most significant contributor to the TEDE in stainless and i carbon steel components was determined to be Co-60. The calculated radioactivity l [ composition of the activated metallic portions of the PCRV and Internal Components l three years after shutdown is: Fe-55, 85%: Co-60, 3.0%: and H-3,12%. These components will be removed prior to conducting the final survey. \\ \\ The PCRV concrete /rebar mixture contains many activation products due to the presence l of trace elements. In the short term, Co-60 is the dominant beta-gamma emitter, while { Eu-152 and Eu-154 are the dominant long term beta-gamma emitters. The nuclides contributing most to the total activity are Fe-55 and H-3. Approximately 31 inches of the activated concrete nearest to the reactor core will be removed prior to conducting the { final survey. 7 Due to the unique design of the FSV HTGR primary and secondary systems, there was l limited potential for migration of activation products and radioactive contamination to r i vtw94 3-4 REVISION 0

b FSV FINAL SURVEY PLAN FINAL SURVEY OVERVIEW L, surfaces outside of the PCRV. This conclusion is supported by the contamination survey results reported in the Fort St. Vrain Initial Radiological Site Characterization Report [ submitted to the NRC in 1991. Ten samples were collected for extensive, independent laboratory analysis of HTDN during the Site Characterization Survey from various levels of the Reactor Building and from the interior surface of the Radioactive Gas Waste System. During normal power operations, the Radioactive Gas Waste System collected gases from contaminated systems throughout the plant. The ratio of radionuclides in the contamination in these samples was found to differ significantly from the ratio found in the activated portions of the PCRV and Internal Components. Fe-55 was found to be the most significant contributor to the activity found in the samples collected outside of the PCRV. The ratio of Fe-55 to Co-60 in these samples was 2.2 as determined in November,1991. Accounting for the radioactive decay of Fe-55 and Co-60 through December,1995 will result in a ratio of 1.3. Draft NUREG/CR-5849, " Manual for Conducting Radiological Surveys in Support of License Termination," June 1992 (Ref. 9) suggests that it may be appropriate for sites { with multiple radionuclides at the time of license termination to account for those radionuclides which would contribute greater than 10% of the total radiation dose from all contaminants or which are present at concentrations which exceed 10% of their ( respective guideline values. Smear samples were collected to identify the major contaminants in accordance with 10 CFR 61 for the 1993 Dry Active Waste Stream originating from decommissioning activities associated with the PCRV. Decay-corrected ( through December,1995, the composition included: Fe-55, 74.2 %; H-3,10.9 %; Co-60, 8.6%; and C-14,1.0%. The balance of the activity, less than 7%, was contributed by nuclides contributing less than 1% each. Of these contaminants, Fe-55, H-3, and C-14 are considered as HTDN since they are not beta-gamma emitters and are not readily detected with most conventional field survey instruments. It is possible that the concentrations of these nuclides could exceed 10% of their respective contamination guideline values. However, due to their significantly smaller dose equivalent factors, their contributions to the total effective dose equivalent would not exceed 10% of the radiation dose from all contaminants. A FSV site-specific method for evaluation, of HTDN sample results has been developed and is discussed in Section 5.2.4. 3.3 Determination of Site-Specific Limits The initial FSV site characterization survey, PCRV concrete characterization and dry [ active waste samples have indicated the presence of multiple radionuclides, including HTDN. Therefore, surface contamination limits have been established using the guidance provided in Regulatory Guide 1.86, and site-specific limits have been developed to account for HTDN The limits for HTDN were submitted to the NRC for approval by letter dated December 23,1993, " Final Radiation Survey Plan, Treatment of Hard to Detect Nuclides," (Ref.10). .) l Prior to issuance of the referenced draft guidance, application of Regulatory Guide 1.86 { contamination limits involved measurement of only those readily-detected beta-gamma and alpha-emitting contaminants as may be present. Evaluation of the radionuclide r 2/17/94 3-5 REVISION 0

i FSV FINAL SURVEY PLAN FINAL SURVEY OVERVIEW b. inventory at FSV has shown that adjustment of readily detectable radionuclide surface contamination limits to account for HTDN cannot bejustified by As-Low-As-Reasonably-Achievable (ALARA) considerations (due to resulting increases in survey equipment and manpower requirements) and would not provide a measurable reduction in the risk to public health and safety. Therefore, the following site-specific limits are proposed for use at FSV. 3.3.1 Alpha and Beta-gamma Removable Surface Contamination Limits The levels of readily-detected removable alpha and beta-gamma surface contamination will be limited to the applicable values provided in Table 3.1. No fuel failures were experienced in the past and no reactor-produced transuranic alpha emitters were identified outside the PCRV by surveys performed during plant operation or in the Initial Radiological Site Characterization. However 5% of the smear samples collected during final survey of the PCRV, Refuel Floor or fuel handling equipment will be analyzed for unidentified (gross) alpha contamination. Administrative action levels have been established in the event that significant removable alpha contamination is detected. These action levels will be based on the appropriate alpha contamination limits from Table 3.1. If removable alpha contamination is detected, additional investigation and survey measurement using approved survey methods will be performed to determine the c [ area classification. Figure 5.3 provides additional detail for these activities. 3.3.2 Alpha and Beta-gamma Total Surface Contamination Limits r L The levels of readily-detected alpha and beta-gamma total surface contamination (fixed plus removable) will be limited to the applicable values provided in L Table 3.1. 3.3.3 Gamma Exposure Rate Limit r m While no formal criteria exist that establish an acceptable level for gamma ~ exposure rate, the NRC has provided interim guidance which directs FSV to use a limit of 5 pIUhr above background for reactor-generated gamma emitting ~ isotopes as a limiting level for direct exposure from " residual" radioactivity (Ref. 11). This recommended limit is also consistent with statements within NUREG-0586. Based upon this guidance, 5 R/hr above background, averaged over 2 2 10 m for facility buildings and 100 m for outdoor areas, is establi-hed as the gamma exposure rate limit for FSV. The gamma background for outdoor and indoor structural areas will be determined as discussed in Section 4.4.3.f. The gamma exposure rate' for all radionuclides will be measured at a distance of one meter from accessible surfaces in facility buildings and outdoor areas. These measurements will be compared to the gamma background for that surwy unit using the methodology discussed in Section 5.2.2. 2/17/94 3-6 REVISION 0

b FSV FINAL SURVEY PLAN FINAL SURVEY OVERVIEW I. 3.3.4 Hard To Detect Nuclide Limits The following steps will be performed as guidelines for the evaluation of HTDN: Representative samples will be taken from various surfaces, structures and systems during or after conclusion of decommissioning activities and analyzed to identify the average HTDN activity due to nuclides such as Fe-55, C-14 and H-3. Numbers of samples and scaling methodology comparable to those typically used for 10 CFR 61 waste characterization will be used to identify and quantify HTDN. The ratios of HTDN to readily detected radionuclides will be compared to previously established ratios. Where differences are observed which could { indicate a significant increase in contribution to the annual TEDE, scenario analyses will be performed to determine the added contribution to the annual TEDE from HTDN. I The annual TEDE due to the presence of HTDN will be detennined using the methods provided in NUREG/CR-5512, Volume 1 (Ref.12). In cases where HTDN dose equivalent does not exceed 2 mrem per year, no additional actions will be taken with respect to remediation. If the dose equivalent due to HTDN is greater than 2 mrem per year, further evaluation l will be performed to determine the TEDE for HTDN plus average concentrations for all other contaminants. If the TEDE for the scenario does ( not exceed 10 mrem per year, no additional remediation will be required j solely for the remediation of HTDN. I 3.3.5 Soil and Water Activity Limits The TEDE for the average concentrations of radioactive materials above background in soil and water will be detennined in accordance with the methodology contained in NUREG/CR-5512, Volume 1 to assure that the total { effective dose equivalent could not exceed 10 mrem during a period of one year. 3.3.6 Administrative Action Levels b Administrative action levels have been established identifying additional investigative actions in the event that measurable contamination is detected at levels below Reg. Guide 1.86 limits. Administrative action levels include: Level 1: Removable alpha or beta-gamma contamination in excess of 25% of 2 the site-specific limit (e.g., beta-gamma >250 dpm/100 cm ), [ [ 2/17/94 3-7 REVISION 0

• FSV FINAL SURVEY PLAN FINAL SURVEY OVERVIEW Level 2: Average alpha or beta-gamma total surface contamination (fixed plus I

removable) in excess of 50% of the site-specific limit (e.g., beta-2 gamma >2500 dpm/100 cm ), Level 3: Average alpha or beta-gamma total surface contamination (fixed plus removable) in excess of 75% of the site-specific limit (e.g., beta-2 gamma > 3750 dpm/100 cm ), Level 4: Maximum alpha or beta-gamma surface contamination (fixed plus I removable) in excess of 75% of the site-specific limit (e.g., beta-2 gamma > 11,250 dpm/100 cm ), l 12 vel 5: Removable alpha or beta-gamma contamination greater than the site-2 specific limit (e.g., beta-garnma 21000 dpm/100 cm ), l Level 6: Average alpha or beta-gamma total surface contamination (fixed plus removable) greater than the site-specific limit (e.g., beta-gamma 2 > 5000 dpm/100 cm ). Level 7: Individual gamma exposure rates greater than 10 pR/hr above the natural background and greater than the standard deviation of the I survey unit set of measurements at the 99.9% confidence level. I lxvel 8: Gamma exposure rates greater than 5 R/hr above the natural background. I Level 9: Calculated annual TEDE in excess of 10 mrem from licensed material in soil or water. Where numerically smaller than the FSV release limits, these Administrative action levels will initiate activities to ensure that sufficient additional survey data is collected to adequately characterize survey areas for either release or additional l remedial measures. In no instance are the Administrative action levels intended to replace the established FSV release limits for determining the suitability for unrestricted use. The actions to be taken upon exceeding these levels are discussed in Sections 4.0 and 5.0. 3.4 Organization and Responsibilitics Final Survey Plan development and implementation will be performed by qualified I members of the Westinghouse Team, primarily the SEG Radiation Protection Department. Position descriptions delineating responsibilities and interfaces are contained in the Decommissioning Plan (DP) and in the FSV Decommissioning Project Radiation Protection Manual (Ref.13). i I L~ 2/17/94 3-8 REVISION 0 l

' FSV VENAL SURVEY PLAN FINAL SURVEY OVERVIEW 3.5 Quality Assurance r-L 3.5.1 General Provisions The Decommissioning Project Quality Plan (PQP) is based on the requirements (' of 10 CFR 50, Appendix B as they tpply to decommissioning activities. The PQP identifies the Final Survey Quality Assurance Plan which will be {- implemented during performance of the Final Site Survey. The Final Survey Quality Assurance Plan identifies the applicable criteria of the NRC approved SEG Quality Assurance Program and the extent to which they apply to implementation of the Final Survey Plan. Examples of the QA Program application are described in the subsections that follow. Additionally, the FSV Decommissioning Project Radiation Protection Program incorporates an internal quality control verification process which is described in this section.

a. Selection and Training of Personnel Qualification requirements and responsibilities have been established for key personnel conducting the final survey and are controlled by the FSV Decommissioning Project Radiation Protection Manual.

A training and qualification program will be established for technician and supervisory personnel selected to conduct the final survey and will include formal classroom training and performance-based training. Training and qualification records will be maintained for all personnel.

b. Instrumentation Selection, Calibration and Operation Selection and use of instrumentation will ensure sensitivities will be sufficient

( to detect radionuclides at the minimum detection requirements specified in Section 4.2.2. Furthermore, the selection and use of instrumentation will assure the validity of the survey data. The selection of this instmmentation has been made based on lessons learned from similar decommissioning and Final Survey Projects. Instrument calibration will be performed either under approved procedures within the FSV Decommissioning Project Radiation Protection Manual using calibration sources traceable to the National Institute of Standards and Technology (NIST), or by qualified vendors with the results ( traceable to NIST. Measurements will be performed using approved written procedures for each instrument. Issue, control, and operation of all survey instrumentation will be established by instrumentation control procedures. y

c. Survey Documentation

( Records of final surveys will be maintained in survey packages according to procedures contained in the FSV Decommissioning Project Radiation Protection Manual. A separate package will be prepared for each survey [ The survey package will be the primary method of controlling and area. I 2/17/94 3-9 REVISION 0 {

FSV VINAL SURVEY PLAN FINAL SURVEY OVERVIEW ' tracking the hard copy records of final survey results. The typical records compiled in a survey package are shown below: Final Survey Package Worksheet Final Survey Comment Addendum (as needed) [ Survey Unit Diagram Random Sample sheet Photographs of the survey area [!- Printout of smear survey analyses Printout of gamma spectroscopy results (if performed) [ Printout of direct survey raw data results Documentation of changes or corrections to survey data Result of HTDN Evaluation (if performed) Survey Package Review Sheet The survey packages will be tracked using a database for survey progress, status and completion, and will be maintained in accordance with appropriate-record-keeping procedures. (

d. Quality Control - Verification Sampling Verification field and laboratory measurements will be performed

( independently on selected samples of survey measurements on an ongoing basis. Instruction regarding the type and number of Quality Control measurements will be contained within the related survey package. The [ results of these analyses will be evaluated to establish the precision and accuracy of the final survey data.

e. Written Procedures All final survey tasks which are essential to survey data quality will be implemented and controlled by procedures contained in the FSV Deconunissioning Project Radiation Protection Manual.

f. Verification of Procedures and Processes 4 [ Verification of the procedures and processes will be performed in accordance with requirements for technical verification of procedures contained within the FSV Decommissioning Project Radiation Protection Manual.

g. Chain of Custody 1

i h Procedures contained within the FSV Decommissioning Project Radiochemistry and Radiation Protection Manuals will establish responsibility for custody of samples and survey data between the point of measurement or -l [ collection until final results are obtained. Final survey sample accountability [ 2/17/94 3-10 REVISION 0

[ FSV FINAL SURVEY PLAN FINAL SURVEY OVERVIEW { will be maintained from the initial sampling until disposal of the sample. This will be accomplished by assigning initial sample custody to the individual [' obtaining the sample. Samples will be maintained in the possession of the individual, under direct surveillance or secured. When custody is transferred (e.g., when samples are sent off-site to another lab for analysis), a Chain of (. Custody Record will accompany the sample for tracking purposes. Chain of Custody Records will be maintained in the records management system.

h. Records Management

{ Generation, handling and storage of the original final survey design and data packages will be controlled by an approved procedure contained within the FSV Decommissioning Project Radiation Protection Manual.

i. Independent Review of Survey Results The survey package from each survey unit will be given independent review to verify all documentation is complete and correct, and that release limits have been met. Also, an Independent Third Party has been selected who will perform a confirmation survey upon completion of the final survey,

j. Control of Surveyed Areas and Systems Administrative (i.e., procedural) and physical controls for survey units will be established to preclude the possibility of contamination subsequent to

[ completion of the final survey.

k. Control of Vendor Supplied Services Quality-related services, such as instrument calibration and laboratory sample analysis, will be procured only from qualified vendors whose internal Quality Assurance Program is subject to audit in accordance with the FSV Project Quality Plan.

3.5.2 Final Survey Quality Control Procedure [ A Final Survey procedure contained in the FSV Decommissioning Project Radiation Protection Manual will establish the quality activities not addressed in other procedures. These activities may include: Conduct of verification sampling measurements Verification of survey measurement data I [ Testing of computer calculations Documentation of surveys Custody of instmments, samples and measurement data [ 2/17/94 3-11 REVISION 0 ______________._m

FSV FINAL SURVEY PLAN FINAL SURVEY OVERVIEW ~ 3.6 Training To the maximum exient possible, personnel assigned to implement the Final Survey Plan will be selected from SEG Radiation Protection Department personnel currently engaged { in deconunissioning and dismantlement phases of the project. Minimum qualincations and prerequisite training have been established within the {- framework of the existing FSV Decommissioning Project Radiation Protection Manual and in accordance with the DP. ( Before assignment to Final Survey, personnel will receive additional formal training including: ( Overview of the Final Survey Plan Objectives of final survey Procedures governing conduct of the Gnal survey [~ Operation of the appropriate field and laboratory instrumentation used in the final survey Performance-based training in the collection of final survey measurements and I samples Records of training, including testing to demonstrate qualification, will be maintained in accordance with established procedures. l 3.7 Laboratory Services The FSV Radiochemistry Program is currently being conducted in accordance with an [ approved QA Program. On-site laboratory radioanalytical capabilities will be utilized to the maximum extent possible to support the final survey. On-site capabilities include gamma spectroscopy (HPGe) of smear and bulk samples, low energy beta /x-ray ( spectroscopy, liquid scintillation, and gas proportional counting. Off-site laboratory sample analysis will be procured only from qualified vendors whose internal QA Program i is subject to audit in accordance with the Project Quality Plan. Contracts are currently ( in place with two qualified vendors for specialized radiological analysis of samples on an as-needed basis. I 3.8 General Survey Plan { Three categories or types of survey areas have been established including: Structures, including building interiors and exteriors { Plant systems, including process piping or ventilation Outdoor areas, including facility grounds and liquid effluent pathway 2/17/94 3-12 REVISION 0

FSV FINAL SURVEY PLAN FINAL SURVEY OVERVIEW These categories have been selected in order to group similar physical characteristics into survey units. The survey effort for each survey unit will be based upon its classification as affected or unaffected as discussed in Section 3.8.2. Due to the large scope of the final survey and the requirement that some survey activities b be conducted in parallel with decommissioning work, a systematic approach is necessary. Further, it is essential that key interfaces between survey activities ~ and other [ decommissioning work activities be identified. t The final survey planning and implementation process for each survey unit will involve the following activities: History file preparation Initial classification Walkdown Turnover for final survey Survey design Modifications Work planning and scheduling Survey instructions i Field support Survey measurements Restoration and isolation .l These steps are described in the following sections. 3.8.1 History File Preparation The history file will be a compilation of relevant operational and decommissioning j data in a standardized format. The purpose of this process is to provide a substantive j { basis for the survey unit classification, and hence the level of intensity of the final 1 survey. Similarly, for structures and outdoor areas, the extent of radioactive I materials involvement in the area (if any) is summarized. The history file may -l contain: Review of the system description Operating history which could affect radiological status The FSV Initial Radiological Site Characterization Report Radiological surveys performed during decommissioning phases ( Other relevant information 3.8.2 Initial Classification The initial classification of the survey units into "affected" and " unaffected" areas provides an overall planning basis for the final survey. All areas of the FSV site f-(including structures, plant systems and outdoor areas) will not have the same 2/17/94 3-13 REVISION 0

I FSV FINAL SURVEY PLAN FINAL SURVEY OVERVIEW potential for residual contamination and therefore do not require the same level of survey coverage to achieve an acceptable level of confidence that an area satisfies the established release limits. Affected and unaffected areas are defined as follows: b Affected areas: Areas that have potential radioactive contamination (based on plant operating history) or known radioactive contamination (based on [ past or preliminary radiological surveys). Affected areas are further L subdivided, according to survey effort, into suspect and non-suspect areas. Unaffected areas: All areas not classified as affected. These areas are not { expected to contain residual radioactivity, based on a knowledge of site history and previous survey information. Survey data contained in the FSV Initial Radiological Site Characterization Report and assessment of Westinghouse Team and PSC personnel knowledgeable of the [ facility conditions will be utilized when applying the classification criteria contained in Draft NUREG/CR-5849 and in the final survey procedures. 3.8.3 Walkdown The walkdown will be a key activity in the preparation of the survey design. For systems, it will include review of system flow diagrams, piping drawings and a physical walkdown of the system. Structures and outdoor areas will also be walked down. The principal objective is to assess the physical scope of the-survey area and to identify the breakdown into units and subunits. Specific requirements will be identified for accessing the survey area and support functions necessary to conduct the final surveys, such as scaffolding, component disassembly, interference removal, engineering modifications, electrical tagout and system alignment to provide access for surveys. Safety concerns such as { access to confined spaces, high walls, and ceilings will be identified. It is noted that for survey units involved with decommissioning activities, the walkdown is best completed when the final configuration is known, usually near or after ~ completion of decommissioning activities. 3.8.4 Turnover for Final Survey Prior to acceptance of a survey area for the final survey, a number of conditions will be satisfied. These include: Decommissioning activities having the potential to contaminate the survey unit must be completed. [ All tools and equipment not required to perform the survey must be removed. Ilousekeeping and area cleanup must be completed. 2/17/94 3-14 REVISION 0

FSV FINAL SURVEY PLAN FINAL SURVEY OVERVIEW Decontamination of affected structural areas and system components must be completed. Operational radiological surveys have been performed. A physical walkdown prior to turnover will be performed to ensure the above [ conditions have been met. Scaffolding needed to be left in place for conducting - the final survey should be identified during turnover preparation. Results of the [ operational radiological survey verifying the status of the area will be included L in the final survey package. { 3.8.5 Survey Design Establishing a uniform grid system is one of the methods that can assist in the j { systematic selection of measuring and sampling locations while providing a mechanism for referencing a measurement to specific locations. Grids are also convenient means to determine average activity levels. b A salient feature of the FSV Final Survey Plan is to minimize (limit) establishing a uniform grid system without compromising the quality of survey data and ensuring the following: 1 I Reference of radiological survey data to a specific survey location in order to l evaluate significance of elevated activity levels (hot spots) Correlation of radiological survey data to a specific surface of known area in o:cer to estimate average activity levels Collection of an adequate number of measurements or samples from within j a designated surface area to enable statistical evaluation I l The methods used will provide sound statistical conclusions for determination of the final survey status. Additional detail is provided in Section 4.0. 3.8.6 Modifications After the survey design is prepared, component disassembly or system [ modification requirements _will be identified. In instances' where disassembly or modifications are required for surveys, a work control process will be used to coordinate activities. This process will include prerequisites for the protection of systems and structures from potential contamination after completion of the survey. 2/17/94 3-15 REVISION 0 h

VSV VINAL SURVEY PLAN FINAL SURVEY OVFRVIEW 3.8.7 Work Planning nnd Scheduling h Upon completion of any required engineering review, the physical modifications will be specified. Field work will be implemented via a work control process described in documented procedures. The process will identify components which [' require opening, identify modifications, indicate restoration requirements and indicate whether a system is to be isolated or returned to service. The process [ will also be used to initiate support work and tagouts necessary for surveys of stmetural and outdoor survey units. The survey unit support work will then be placed on the work schedule for implementation. 3.8.8 Survey Instructions [ The final survey instmetions will be containe<1 within the survey package. These instructions will specify the number and type of radiological measurements to be taken at each location or component identified in the survey design. The [ instmetions will also identify other samples to be collected. The survey I instructions will be prepared in accordance with procedures contained in the FSV Decommissioning Project Radiation Protection Manual. 3.8.9 Vield Support The work control process will identify each component or survey. location requiring support work and tagouts. In cases where special surveys are required such as components, embedded piping, or large tanks which are classified as affected, other preparation work may be required. This may include gridding of j sumps once access is provided and safety precautions have been satisfied. 3.8.10 Survey Measurements Final survey measurements will be conducted in accordance with procedures contained in the FSV Decommissioning Project Radiation Protection Manual and specific survey instructions for the survey unit provided in the survey package. A sufficient number of measurements will be taken to conclusively demonstrate that the release limits have been met. The measurements will be obtained by conducting surveys using approved methods and techniques such as surface scans, direct measurements of surface contamination, smear samples for removable surface contamination. and gamma exposure rate measurements. These measurements are discussed in Section 4.0. 2/17/94 3-16 REVISION 0 e

i FSV FINAL SURVEY PLAN FINAL SURVEY OVERVIEW 3.8.11 Restoration and Isolation

a. Systems After final survey measurements have been taken, Icviewed and approved, and any necessary verification survey measurements have been completed, the system will be restored and components replaced as specified in the work

[ control process. If required, the system will be isolated to protect against intrusion of radioactive contamination. Isolation and control of plant systems after completion of the final survey will be performed under an approved { procedure. Many plant support and service systems will be returned to service after { completion of final survey measurements. Examples are: compressed air, heating and cooling, ventilation and fire protection. When a system is required to be returned to service following completion of the final survey, [ administrative controls and physical barriers will be used to minimize the possibility of system contamination. These include, but are not limited to, locked access and surveillance activities to ensure that the system is not [ aligned or operated in a manner which could compromise the integrity of the final survey results.

b. Structures and Outdoor Areas After measurements are completed in structures and outdoor areas, controls will be used to prevent or minimize possible radioactive contamination.

These will be controlled by a procedure, as appropriate. Numerous structural survey units have been defined within the Radiological Controlled -Area l [ (RCA) of the FSV facility. These include all of the structural survey units which comprise the Reactor Building, portions of the Turbine Building, the { Radwaste Compactor Building, the New Fuel Storage Building, Technical Support Building which contains the Radiochemistry Laboratory, and the Liquid Effluent System and Liquid Effluent Discharge Pathways. As surveys l { are completed in contiguous RCA survey units (with completion of the surveys being indicated by completed, approved release records), reduction i of the RCA boundaries will be made. Removal of such areas from the RCA [ provides additional assurance that material containing radioactivity is not used in, or transponed through previously surveyed areas. 3.9 Schedule The final survey is plarmed to be completed in four major phases which encompass distinct portions of the facility. The first phase is expected to include unaffected outside area survey units, unaffected structural survey units outside the RCA, and unaffected systems and structural survey units within the Turbine Building. The second phase is expected to include unaffected systems and structural survey units within the Reactor 2/17/94 3-17 REVISION 0 m

[ l FSV VINAL SURVEY PLAN FINA:. UJRVEY OVERVIEW [. Building. The third phase is expected to include affected systems and the remaining survey units affected by the removal and temporary storage of irradiated and/or l (- contaminated components from the facility. The final phase is expected to include laydown areas and temporary radioactive waste storage areas, and the liquid effluent j pathways. Upon completion of each phase, the release records will be compiled and the I survey units will be made available for NRC Confirmatory Surveys. [ 3.10 Final Survey Reporting Interim Reports as well as a Final Summary Report will be prepared for submittal to the ( NRC to meet the intent of Regulatory Guide 1.86 for final survey reporting. The repons will follow the guidance of Draft NUREG/CR-5849 regarding content and fonnat. Interim Reports will be submitted upon completion of each of the major survey phases ( discussed in Section 3.9 above. A Final Summary Repon will be prepared summarizing results of the entire decommissioning project. The Interim and Final Summary Reports are described in Section 6.0, " Final Survey Reporting" l L F L i I l r u W-W 2!!7/94 3-18 REVISION 0 __-_.J

FSV FINAL SURVEY PLAN SURVEY: PLAN AND PROCEDURES 4.0 SURVEY PLAN AND PROCEDURES 4.1 General [ The design approach of the FSV Final Survey reflects the final configuration of the facility, which will largely be left intact with the majority of equipment abandoned in { place. The major survey effort will be confined to the survey units contained within the Reactor Building, portions of the Turbine Building where radioactive materials were used or handled, the Radwaste Compactor Building, the New Fuel Storage Building, the ( Technical Support Building which contains the Radiochemistry Laboratory, the Liquid Effluent System, Liquid Effluent Discharge Pathways, and remaining portions of the l plant systems identified as contaminated or potentially contaminated in the FSV Initial h Radiological Site Characterization Repon. These survey units will be classified as affected. If radioactive material process or storage areas are established outside these survey units to support decommissioning operations, these areas will be identified and classified on a case-by-case basis. The remainder of the areas within the scope of the survey will be classified as J unaffected. All radioactive material handling, movement and storage on the site has been i controlled under approved procedures. No significant FSV actions have resulted in a j radiological impact on the site envimns as evidenced by extensive site surface measurements and soil sampling reported in the FSV ' Initial Radiological Site Characterization Report, the Radiological Environmental Monitoring Program (REMP), the Environmental Radiological Surveillance Program (ERSP) and the study conducted by Colorado State University's Department of Radiology and Radiation Biology from June 1990 until October 1990. The later study was used to assess the pre-operational radiological environmental status of the Independent Spent Fuel Storage Facility (ISFSI) ' site. In addition, procedures have been established to document events that could result in a radiological impact to the site during decommissioning. The Final Survey Plan and implementing procedures are designed to focus primarily on remaining plant structures and systems in the affected areas. Instrumentation has been selected and survey procedures developed to detect and measure surface contamination levels (primarily Co-60 and Eu-152) and gamma exposure levels in these affected areas. I-4.2 Instrumentation ] l Radiation detection and measurement instrumentation for final surveys has been selected j to provide reliable operation and adequate sensitivity to demonstrate that' the 'i measurements taken are sufficient to conclusively demonstrate that the release limits have been met. Commercially available portable and laboratory instmments and detectors { produced by several manufacturers have been selected based upon detection sensitivity,. operating characteristics and expected performance in the field. The detectors selected [ and their detection characteristics are summarized in Tables 4.1 and 4.2. Recording instruments (survey meters) for use with these detectors have also been evaluated. 2/17/94 4-1 REVISION 9

FSV FINAL SURVEY PLAN SURVEY PIAN AND PROCEDURES [ Instrumentation to be used for gamma exposure rate measurements and special purpose measurements are also described. 4.2.1 Instrument Detector Description The principal detectors selected for final survey measurements are shown in Table 4.1, " Final Survey Instrumentation Summary". The detectors used for total [ surface contamination monitoring will, for the most part, be operated with computer-based data logging survey meters. { 4.2.2 Detection Sensitivity i The detection sensitivity of the detectors selected for field surface contamination { measurements has been evaluated. These results are illustrated in Table 4.2, " Typical Detection Sensitivities for Tc-99 and Th-230". The results are shown for the principal instruments that are expected to be used for field beta-gamma j [ and alpha total surface contamination measurements. The minimum detectable j activities (MDAs) for field equipment are calculated by using the SEG proprietary spreadsheet program, CNTTIME* (Ref.14), an MDA and limit value software. Laboratory instrument minimum detectable activities are calculated using Equation 4.1. The MDA is dependent upon several-factors: sample count time, background count time, background count rate and detector efficiency. Smear counters for measurement of removable surface contamination are of modern design (anti-coincidence low-background) and will be used to determine the [ activities for both alpha and beta radiation. The gamma exposure rate measurement instmment sensitivities for field equipment in the normal modes of operation are: Pressurized ion chamber (Reuter-Stokes): 1-3 pR/hr { Ludlum Model 44-2/ sodium iodide detector: 2-3 R/hr. 2/17/94 4-2 REVISION 0

FSV FINAL SURVEY PLAN SURVEY PLAN AND PROCEDURES TABLE 4.1 FINAL SURVEY INSTRUMENTATION

SUMMARY

Radiation Detector Type Detector Manufacturer Units Detected Area-Density & Model # Surface Gas flow 125 cm2 Ludlum cpm 2 Beta-Gamma, proportional 0.8 mg/cm 43-68 Alpha 2 Surface Gas flow 300 cm Ludlum cpm 2 Beta-Gamma, proportional 0.8 mg/cm 43-47 Alpha l Surface Gas flow 550 cm Ludlum cpm i 2 Beta-Gamma, proportional 0.8rra em: 43-37 Alpha 2 Surface ZnS(Ag) Plastic 125 cm L adium cpm [ 2 Betr Alpha Scintillation 1.2 mg/cm 43-89 2 krface Geiger-15.5 cm Ludlum cpm 2 Beta. Gamma Mueller 1.7 mg/cm 44-40 Surface & Liquid Liquid N/A Beckman cpm Contamination Scintillation LS 3801 Beta /Lray Removable Gas flow 24.2 cm2 Tennelec cpm 2 Surface proportional 80 pg/cm LB-5100 Beta-Gamma, 1 Alpha f Gamma Pressurized Ion 8 L Sphere Reuter-Stokes pR/hr 2 [ Exposure Rate Chamber (PIC) 2.85 g/cm RSS-112 Ganuna Sodium Iodide 1" X 1" Ludlum R/hr Exposure Rate scintillation 44-2 (Cs-137) Gamma Iligh-purity 2" X 2" Canberra Geometry Spectroscopy Germanium Genie-PCS Dependent

Background

High-purity 2" X 2" EG&G Ortec Geometry and Special germanium Nomad Dependent Measurements L r 2!17/94 4-3 REVISION 0

1 FSV FnNAL SURVEY PLAN SURVEY PLAN AND PROCEDURES TABLE 4.2 i TYPICAL DETECTION SENSITIVITIES FOR Tc-99 AND Th-230 Count Ilackground (. time Ilackground Efficiency

• Count Time MDA Scan Survey 2

Irtstrument Radiation (min) (cpm) (epm /dpm) (min) (dpm/100 cm ) MDA 2 (dpm/100 cm ) L.udlum beta .05 750 0.21 1 1779 2300 Model 43-68 Ludlum alpha 0.17 3 0.20 5 122 570 Model 43-68 Ludlum beta-gamma 0.2 50 0.12 5 3580 5000 Model [ 44-40 (scaler mode) ( Tennelec beta 1 3 27.9 20 20 N/A LB-5100 Telmelec alpha 1 0.3 29.1 20 N/A I LB-5100 i Table 4.2 Notes: [ ' Sources used for efficiencies are traceable to the National Institute of Standards & Technology (NIST). Sources consist 2 2 of Tc-99 or Th-230 uniformly deposited over areas of 17.3 cm or 100 cm. [' The efficiency is determined by counting the source with the detector in a fixed position approximately one-half cm from the source. Scan survey sensitivity (MDA) was calculated assuming a scan rate of 5 cm/second. r L I E L r L r 2/17/94 4-4 REVISION 0

VSV FINAL SURVEY PLAN SURVEY PLAN AND PROCEDURES 4.2.3 Minimum Detectable Activity Calculation ( MDA values for smear counting instrumentation will be calculated as shown below: 2.71 R R 3 3 + 3.29 -+- (4,1) t' ) t* t* l L MDA = E r L where: [ MDA = the minimum amount of activity that can be statistically detected above background with a 95% probability and with a maximum of 5% probability of falsely interpreting background activity as activity due to I contamination (dpm/100 cm ), 2 L t, = sample counting t me (minutes), i t R = the background count rate in counts per minute (cpm), 3 t, = background counting time (minutes), and E = the counting efficiency, (cpm /dpm). 4.2.4 Calibration and Maintenance Instmments and detectors used to conduct final surveys will be calibrated and maintained in accordance with instrumentation procedures contained in the FSV Decommissioning Project Radiation Protection Manual. If vendor services are ~ used, these services will be conducted in accordance with approved procedures and an internal QA Program is subject to audit in accordance with the PQP. Radioactive sources used for the purpose of calibration will be traceable to the National Institute of Standards and Technology (NIST) for both FSV and vendor operations. ] 4.3 Survey Plan 4.3.1 Classification All areas of the FSV site (including stmetures, plant systems and outdoor areas) will not have the same potential for residual contamination and will not require c the same level of survey coverage to achieve an acceptable level of confidence Wat the site satisfies the established release limits. Therefore, each survey area 2n v94 4-5 REVISION 0

FSV FINAL SURVEY PLAN SURVEY PLAN AND PROCEDURES (. will be classiGed as "affected" or " unaffected" in accordance with final survey procedures. Affected and unaffected areas are deGned as follows: unaffected areas: All areas not classified as affected. These areas are not expected to contain residual radioactivity, based on a (- knowledge of site history and previous surve,y information. affected areas: Areas that have potential radioactive contamination (based on plant operating history) or known radioactive contamination (based on past or preliminary radiological [_ surveillance). Affected areas may be further subdivided, according L to survey effort required, into suspect and non-suspect areas as follows. Floors and lower walls (up to two meters above the floor) of affected areas will be designated as suspect and receive 100% i [ survey coverage. Upper walls, ceilings and equipment (more than two meters above the floor) may be designated as suspect or non-suspect depending on the potential for residual activity based on j operating history and previous surveys. These areas may be designated as non-suspect if the average contamination levels do not exceed 50% of the release criterion for total surface contamination or 25% of the release criterion. for removable contamination. Upper walls, ceilings and equipment classified as non-suspect will not require 100% survey coverage but will require [ a sufficient number of measurements (usually a minimum of 30 i measurements each) from horizontal and vertical surfaces to ensure that the mean value is less than the average surface contamination [. limit at the 95% confidence level. Initial classification of individual survey units is based on data provided in the FSV Initial Radiological Site Characterization Report, history of radioactive materials involvement or potential for contamination of the survey unit and [' recommendations of Westinghouse Team and PSC personnel knowledgeable of the facility conditions. Table 4.3, " Initial Facility Survey Area Classification," illustrates the initial classification of facility structures, systems and outside { locations. These classifications will be either verified or updated during conduct i of the final survey. [ Final classification of individual survey units will be performed following the logic depicted in Figure 4.1, " Survey Unit Classification Process." r L f L. 2/17/94 4-6 REVISION 0

FSV FINAL SURVEY PLAN SURVEY PLAN AND PROCEDURES i TABLE 4.3 INITIAL FACILITY SURVEY AREA CLASSIFICATION LOCATION AFFECTED UNAFFECTED INSIDE RESTRICTED AREA Reactor Building 100 % Turbine Building 10 % 90 % [ Search and I.D. Building 100 % Gas Charging Building 100 % Technical Support Building 20 % 80 % Evaporative Cooling Building 100 % Switchyard Relay Building 100 % ) Main Warehouse Building 10 % 90 % j New Fuel Storage Building 100 % [ Building 10 10 % 90 % Flammables Storage Building 100 % {- 3 Security Administration Building 100 % North Warehouse Building 100 % Electrical Warehouse Building 100 % [ Machine Shop Building 10 % 90 % Weld Shop Building 100 % Equipment Building (PSC Carpenter Shop) 100 % Construction Workshop Building 100 % Firewater Pumphouse 100 % { Radioactive Waste Compacting Building 100 % Service Water Pumphouse 100 % ( Insulator Building 100 % Chemical Building 100 % Cire. Water Make-up Pumphouse 100 % [ 2/17/94 4-7 REVISION 0

FSV FINAL SURVEY PLAN SURVEY PLAN ANL< PROCEDURES TABLE 4.3 (CONT.) INITIAL FACILITY SURVEY AREA CLASSIFICATION 4 LOCATION AFFECTED UNAFFECTED i Pole Barn Storage Building 100 % QA/QC Lab. and Snubber Building 100 % Kennedy Building 100 % Chicago Pneumatic Building 100 % ) Outside Aux. Boiler Building 100 % ACM Diesel Generator Building 100 % I Outside Temporary Storage /Laydown Areas 100 % OUTSIDE RESTRICTED AREA Employee and Construction Parking Lots 100 % PSC Carpenter Shop (Storage Building) 100 % [ Field Office 100 % Engineering Building 100 % Production Conference Rooms 100 % N R.C. Site Inspector's Office 100 % Record Storage 100 % Visitor Center 100 % Receiving Warehouse 100 % Production Training Complex 100 % Decommissioning Project Offices 100 % b Gardening / Equipment Storage 100 % ) Communication Building 100 % East and West Storage Basins 100 % l [ Sewage Lagoon #1 and #2 100 % East and West Evaporation Ponds 100 % f L East and West Settling Basins 100 % I 2/17/94 4-8 REVISION 0 i

b ~ FSV FINAL SURVEY PLAN SURVEY PLAN AND PROCEDURES TABLE 4,3 (CONT.) INITIAL FACILITY SURVEY AREA CLASSIFICATION LOCATION AFFECTED UNAFFECTED Mechanical Draft Cooling Towers 100 % Liquid Effluent Pathways 80 % 20 % Fann Pond 100 % PLANT SYSTEMS System 22, Secondary Cooling System 100 % f System 25, Liquid Nitrogen System 100 % System 29, Gas Charging System 100 % System 31, Feedwater and Condensate System 100 % System 32, Feedwater IIcater Vents and Drain 100 % System 33, Water Treatment System 100 % ( System 41, Circulating Water System 100 % System 42, Service Water System 100 % System 43, System Injection System 100 % System 44, Domestic Water System 100 % System 45, Fire Protection System 100 % System 48, Alternate Cooling System 100 % System 51, Turbine Generator and Auxiliaries 100 % ( System 52, Main Turbine Steam Supply System 100 % System 53, Extraction Steam System 100 % [ 1 System 54, Turbine Lube Oil Purification 100 % System 75, Turbine Building Vents and Drains, 100 % Control Room Service Building IIVAC System 84, Auxiliary Boiler and IIeating System 100 % System 91, Ilydraulic Power System 100 % 2/17/94 4-9 REVISION 0 s

' FSV FINAL SU'RVEY PLAN SURVEY PLtN AND PROCEDURES ( TABLE 4.3 (CONT.) INITIAL FACILITY SURVEY AREA CLASSIFICATION LOCATION AFFECTED UNAFFECTED System II, PCRV Penetration Seal Leak Collection 100 % System ] I System 13, Fuel IIandling Equipment System 100 % System 14, Fuel Storage Equipment System 100 % System 16 Auxiliary Equipment 100 % System 21, Primary Coolant System and Helium 100 % Circulator Auxiliary System System 23, Helium Purification System 100 % System 24, IIelium Storage System 100 % System 46,-Reactor Plant Cooling Water System 100 % System 47, Purification Cooling Water System 100 % System 55, Turbine Vents and Drains 10 % 90 % System 61, Decontamination System 100 % System 62, Radioactive Liquid Waste System 100 % System 63, Radioactive Gas Waste System 100 % System 72, Reactor Building Drain System 100 % System 73, Reactor Building IIVAC System 100 % System 79, Radiochemistry Laboratory Ventilation 15 % 85 % System 93, Controls and Instrumentation 80 % 20 % F. L l= 2/17/94 4-10 REVISION 0

FSV VINAL SURVEY PLAN SURVEY PLAN AND PROCEDURES f 4.3.2 Reference Locations Reference locations will be clearly identified to provide reproducible survey measurement locations. Whenever it is appropriate and cost effective, gridding such as that illustrated in Figure 4.2, " Reference Grid Layout," will be used. However, the physical grid layout may be substituted with the use of physical and logical facility markers that will remain identifiable after decommissioning is completed. Based upon the HTGR design and the large number of systems and equipment to be abandoned in place, sound engineering judgement will be used to divide a survey area into survey units (or subunits) to support reproducible QC and NRC Confirmatory Survey applications. This will be accomplished through selecting reference locations that identify the survey area bounds such as; horizontal and vertical structural support beams, systems, components, piping runs, physical sheet steel weld seams and concrete pour seams. Actual lengths, widths and { distances will be included to support reproducibility. All surfaces will be included and will be equally likely to be selected for surveys. [ The size of the survey unit will be chosen to assure that the total number of data points and/or the spacing (frequency) of measurements / sampling enables a statistical evaluation of the data collected. Survey units will be selected to have ( common history or other characteristics which are naturally distinguishable from other portions of the site. Survey units will not include both affected and unaffected areas. The maximum survey unit size for building surface areas 2 classified as affected will be limited to approximately 100 m of equivalent surface area. In addition, the use of photographs, structural and actual system drawings or maps will be included to support survey area and survey unk (cr subunit) configurations. This method of area designation will also serve to assist suney ] personnel to define areas of elevated activity requiring additional investigational i surveys and/or remediation. Areas where gridding may not be feasible include, but are not limited to, interior structural surfaces, system piping, portions of [ interior walls within the Reactor Building, main steam and feedwater piping areas and where interferences exist and they are not expected to be removed. 4.3.3 Survey Maps I Survey maps will be used to document the measurement locations. Figure 4.3, " Final Survey Unit Map," illustrates a typical survey map of a structural survey unit for the PCRV to be used in the absence of physical gridding. Maps will be prepared for specific survey areas to identify structures, systems or equipment l which define boundaries of survey areas, units or subunits. Where grid layout cannot be performed and gridding is not practical or cost effective, the 2/17/94 4-11 REVISION 0

b ' FSV FINAL SURVEY PLAN SURVEY PLAN AND PROCEDURES b locator / survey maps will provide the means of reproducing survey measurement locations. 4.3.4 Survey Point Identification [ Survey points will be identified by a unique reference location ID code or number. The numbering convention will be to start at the reference location, usually the extreme southwest corner of a survey unit, and proceed sequentially [' north to east to south for walls and east to north for floors and ceilings labelling each survey point in a continuing sequence. Survey points are numbered in sequence for each subunit similarly as described above. An individual reference f location will have a unique identification code as determined by its survey unit (or subunit) ID and the number of the survey points within that unit. ( 4.3.5 Surface Scans [- Scanning surveys will be performed to screen large areas, to search for areas i above the average release limits and to detect localized areas above the maximum release limit. The scanning methods used (instrument and survey technique) will { be designed to detect 75'/o of the average total surface contamination release 2 limits, (e.g.,3750 dpm/100 cm for beta gamma). 4.3.6 Surface Activity Measurements Surface activity measurements will be taken at measurement locations and frequencies based upon the classification of affected or unaffected areas. Specific l guidance regarding the location and number of measurements will be provided in the survey package. The general set of measurements will consist of direct (total) beta-gamma and removable beta-gamma at each measurement location. In areas and systems identified as alpha affected, direct (total) surface and removable surface alpha measurements will also be taken. If data obtained from scan h surveys can be demonstrated to provide reproducible and statistically acceptable data similar in quality to that obtained from direct measurements, direct measurement for total surface activity will not be required. Direct measurement of surface contamination will not be performed on interior ) [ surfaces of plant systems where the results of decommissioning surveys indicate { l only the presence of HTDN. The evaluation of these surfaces will be based upon l the results of indirect measurements. When surveys indicate that contamination levels above the average contamination release limits may be present, appropriate follow-up investigation and/or ( measuremcms will be performed. Areas of elevated activity within buildings or-structures will be tested to assure that the average surface activity level within a contiguous square meter containing the elevated area is less than the release limit. 2/17/94 4-12 REVISION 0

VSV FINAL SURVEY PLAN SURVEY PMN AND PROCEDURES f 4.3.7 Exposure Rate Measurements Gamma exposure rate measurements will be taken at one meter from accessible surfaces on a systematic basis established for these measurements in structures and outdoor areas. In locations where it is not physically possible to locate an ( instrument one meter from the surface (e.g, inside pipe chases), gamma exposure rate measurements will not be taken. Gamma measurements of surfaces will be made at least 1.5 meters from adjacent perpendicular surfaces or the proximity f will be noted in the survey record. This is intended to offset the contribution to the observed gamma exposure rate of naturally-occurring radioactive nuclides in structural materials. 4.3.8 Soil and Water Sampling L Soil and water samples will be collected in accordance with a final survey procedure. For affected survey areas, such as the liquid effluent pathways, one sample will be collected from each grid intersection (or grid equivalent where grid layout is not possible). For unaffected survey areas, a total of 30 random soil and water samples will be collected. Specific guidance regarding the location and number of soil and water samples will be provided in the survey packages. Additional soil and water samples will be collected if a contamination event or spill occurs, or if survey measurements indicate areas of elevated activity from licensed material. 4.3.9 Special Sampling and Measurements

a. Sampling of Sediment and Loose Material

( Samples of loose paint, dust or other sediment will be collected for analysis i as part of biased sampling and measurements. Such samples may be collected in drain receptacles, sumps, and other catchments in affected areas. All storm drain catchments will be sampled and surveyed. These samples will be analyzed by gamma spectroscopy. ~'

h. Embedded Piping Surveys

[ Specialized measurements will be made to demonstrate that normally inaccessible piping (e.g., embedded piping) is below the release limits for surface contamination. This will include but not be limited to the use of [ calibrated detectors extended into piping mns in a controlled manner and the use of data acquisition equipment to document the measurements, f When surveys performed during dismantlement indicate the presence of only IITDN (e.g., H-3), indirect measurements will be used to assess radiological status and suitability for unrestricted use. 2/17/94 4-13 REVISION 0

' FSV FINAL SURVEY PLAN SURVEY PLtN AND PROCEDURES 4.3.10 Sampling for IIard To Detect Nuclides ( Samples will be taken and analyzed from random survey locations in affected areas of the facility, such as portions of the liquid effluent pathway, reactor building, radwaste areas, plant system interior surfaces and the remaining PCRV (' compared to previously established ratios. Where differences are observed which structure. The ratios of IITDN to readily detected radionuclides will be could indicate a significant increase in contribution to the annual TEDE, scenario [ analyses will be performed to determine the added contribution to the annual TEDE from IITDN. 4 [ To limit the number of additional samples, while achieving a statistically sound degree of confidence that surfaces meet release limits, action levels will be l p developed for initiation of additional sampling. Examples of action levels to I t initiate additional sampling are: j Total (fixed plus removable) contamination measured in an affected survey unit that exceeds the release limits, Removable contamination measured in an affected survey unit that exceeds the release limits. 4.4 Background Level Determination i l 4.4.1 General Requirements j l 1 Background response will be established for each type of instrument to be used j for surface contamination and gamma exposure rate measurements. Gamma q exposure rate measurements require determination of the gamma background 1 response of detectors at one meter from surfaces. The relative background I responses of the pressurized ion chamber and micro-R instruments will be determined, In addition, backgrounds will be determined for specialized detectors and detector systems. These include: large area detectors and detectors for in-situ monitoring (GM, Gas-flow Proportional, and NaI (TI) scintillator). 4.4.2 Objectives of Background Determination I '!he objectives of background determinations for FSV final survey measurements are to: f Assure reliable instrument operation; L Establish the reference background values for each type ofinstrument-detector + to be used in the survey; I L i 2/t7/94 4-14 REVISION 0 l .-_.___._._.___m_. __._u .______.m_____m.._

FSV FINAL SURVEY PLAN SURVEY PLAN AND PROCEDURES ( Assess the variability in background responses for principal detectors under different applications and conditions of use; and Determine the need for correction factors or special measurements to establish the background value for Gnal survey measurements. ( 4.4.3 Background Measurements ( Locations will be identified as needed, to acquire background measurement data for each type of measurement. Collection of background data will be performed in accordance with approved procedures appropriate for the instrumentation used. Background determinations for each type of final survey measurement will be performed as described below, a. Direct Surface Beta-Gamma Measurements To determine background for direct surface beta-gamma measurement, data will be collected at the time of survey for each survey unit. The number of background locations is dependent upon the survey unit size and complexity. The counts will be accumulated by a scaler in the preset time accumulation mode. Generally, this measurement will be made in the vicinity of the survey unit using a beta absorber suf0cient to stop all survey unit beta radiation from being measured.

b. Direct Surface Alpha Measurements Similar protocols to those used to determine the direct surface beta-gamma background are used to determine the direct surface alpha background.

Special counting techniques may be required to assess the influence of naturally occurring radionuclides. Background counting times may be accumulated and mean backgrounds used because of the low numbers of . counts observed. If radon is a problem, the detector will be equilibrated to j the radon atmosphere before doing the alpha background and survey. j

c. Removable Surface Beta-Gamma Measurements

[- Background determinations of beta-gamma smear counters will be made by taking a series of measurements of blank smear media or-of a series of smears collected from an area verified free of licensed material. b

d. Removable Surface Alpha Measurements

[ Background determination for removable surface alpha measurement uses the ] same protocols as that for removable surface beta-gamma background. i b 2/17/94 4-15 REVISION 0

FSV VINAL SURVEY PLAN SURVEY PLAN AND PROCEDURES

e. Soil and Water Activity Measurements

( Soil and water samples will be collected from the environs within areas unaffected by licensed operations in order to establish the environmental background for activity in soil and water. The samples collected will include (' soil samples from open land areas, and water and stream sediment samples from upstream locations in the South Platte and St. Vrain rivers. For background soil and water samples where the activity for nuclides of interest exceeds 10% of the release limit at the 95% confidence level, the adequacy of the sample population will be evaluated in accordance with Equation 4.2.

f. Gamma Exposure Rate Measurements Outdoor gamma exposure rate reference values have been previously established and reported within the 1989 Radiological Environmental Monitoring Program Report (15.8 pR/hr) (Ref.15), the 1990 FSV Site Characterization Study conducted by Colorado State University (13.3 pR/hr)

(Ref.16), the 1990 Colorado State University ISFSI Site Background Radiation Study (14.2 pR/hr), the 1990 Radiological Environmental ( Monitoring Report (16.25 R/hr), and the 1991 Fort St. Vrain Initial Radiological Site Characterization Report (15-20 R/hr). These values will be either updated or confirmed by the outdoor background gamma exposure [ rate determination for final surveys. Experience with previous Final Survey projects has shown that the above h values are generally inappropriate for use as indoor background gamma exposure rates. For indoor surveys, there is a large and variable influence on the local background due to the geometry of the natural radioactive materials (e.g., K-40) in the walls and floor of the area being surveyed. In addition, the materials of construction can greatly influence the local gamma background rate by shielding indoor areas from the outdoor gamma background. { Therefore, to demonstrate that the residual gamma radiation from licensed - materials for indoor structural areas does not exceed the limit of 5 R/hr, it is necessary to establish the local gamma background in the area of interest. {- This is done using the set of individual gamma exposure rate measurements of the survey unit to estimate the mean value after having performed the scan and/or direct measurement surveys and the removable contamination survey. [ to verify that there is no residual activity above Administrative Action Levels. Since it will have been demonstrated that there is no residual activity above - Administrative Action Levels, the mean of the set of. individual gamma exposure rate measurements of the survey unit represents the best estimate of the local background for the survey unit. This approach is based on the assumption that the set of measurements represents a normal Gaussian 2/17/94 4-16 REVISION 0

' FSV VINAL SURVEY PLAN SURVEY PLAN AND PROCEDURES distribution. If the set of measurements is observed to be log-normal (i.e., includes outlier data points representative of very localized detectable activity) the median value for the set of measurements is more representative of the best estimate of the local background for the survey unit.

g. Specialized Measurements Detector background is affected when detectors are inserted inside massive components or within piping embedded in concrete. This is particularly I

noticeable when NaI(TI) detectors are used. Thus it will be necessary either to provide a mockup of an embedded pipe, for example, or to develop empirical correction factors for backgrounds when surveying such equipment. I To this end, a series of background measurements will be performed in embedded piping and in large components which can be ascertained to be free of radioactive contamination.

h. Verification of Background Measurement Population l

For background soil and water samples where the activity for nuclides of interest exceeds 10% of the release limit at the 95% confidence level, the population of background measurement will be tested to ensure that the number of measurements in the data set is adequate to support the population statistics. The total number of background measurements required will be calculated as shown below: I I (4.2) n = i a(x) I where: ni = number of data points required I tug = t statistic for 95% confidence at df = n-1 degrees of freedom df = n-1 degrees of freedom where n is the number of initial data points l s = standard deviation of initial measurements l a = variable depending upon background variation, this value is l typically 0.2 l x = mean of mutual determination when greater than, or equal to 10% of the release limit I 4.4.4 Documentation And Control of Background Measurements ] l Background measurements will be collected and recorded in accordance with a j l final survey procedure. 2/17/94 4-17 REVISION 0

~ FSV FINAL SURVEY PLAN SURVEY PLAN AND PROCEDURES b 4.5 Sample Analysis As indicated in Section 3.7, an in-depth sample analysis capability is available on-site to support final surveys. Routine samples of sediment, paint chips and debris will be evaluated using gamma spectroscopy. The need for additional sampling and analysis will be determined on the basis of this initial evaluation. [ F 2/17/94 4-18 REVISION 0

FSV FINAL SURVEY PLAN SURVEY PLAN AND PROCEDURES FIGURE 4.1 Survey Unit Classification Pmcess Survey Unt Boundary Defined if Classify as No Hatory of Rad Yes Unaffected Area Matereits ? Perform biased turnover survey y Perform a 10% Evaluate biased scan survey of surwy measurements floor & lower walls i y Smear and measure beta & Classify as Action f gamma at 30 Non-Suspect Level 1or 2 L random locations Affected Area ? Yes r u l u Scan 100% of Classify as floors a walls <2 m Suspect f Affected Area L-1r Smear and measure S 100% M beta & gamma at g ,,, g systematic or random locations for a minimum of 30 locations on L surfaces (2 m Smear and r measure beta & y q AdministratNe Action Levels Smear and s ema e or Action Level 1 - Removable contamination J U"***' >25% of release limit biased locations on Action Level 2 - Average surface j contamination >SO% of release kmd i i i l 2/17/94 4 19 REVISION 0

FSV FINAL SURVEY PLAN SURVEY PLAN AND PROCEDURES FIGURE 4.2 Reference Grid Layout Survey Unit ID 1 2 3 4 5 6 E....;...;p...;... ;... a;.... 7 D l l l l l l l l l l WALL 1 C....l... . '....:... '....:.... 3091010W01 i 8 ....t,...L...L.. J.... J..... A l l l l l TRENCH 1 [ TRENCH 2 A B C D E E,D,C,B,A [#/[# l l l l 1 ........,...r...,.. C 3 ..... q... q... $,..., 4.. 2 2 B ,...,...c...c.. . ;... q... q... q.... . 4.. 3 l 3 A r-- r-i q 4 WALL 4 WALL 2 1 2 3 4 5 3091010W04 3091010W02 FLOOR ( 3091010F01 5 4 3 2 1 A ...j...;...y...y... g ...,i....i...,l... I... l l l l C ..;...;...;...c... l o 4... 4....;........ E W ALL 3 3091010WO3 LEGEND d..,..- -*.=-mi==. Q tems..rwrser I ( l l A-. IN FILEe l l 1 2/17/94 4-20 REVISION 0

FSV FINAL SURVEY PLAN SURVEY PLiN AND PROCEDURES FIGURE 4.3 Final Survey Unit Map (Example) PCRV SurveyArea sA ..2. -----_..,,. [ n, )/",/ [ yd. tY-I, ' J y-Jp j; 'Q],,J4:M.4JJ / i T I 4'- p,:, i 1 J 1le ,a l[1 Njl n!

j. M~[

.qJ._7 g g

! pc#! p 1>Ht%m7%jj p

is gi ' biii[f 'Q]Jlj;iL.Lg.;;,;48 W t 1d dii i f A ~1T i i l$ 4^ " pJ pt

it M-d,~

p!?e-y:.q r ,+ Lf. saa - +t y

- -.; n '

1. j,;

Ja M.

~p ;

h~e. r e i. p., 4 - r.,,. 9 '- i. F '- na l

• za m

a sza m m / m TendonAnchor a -______l / m / / J-. l $, N.!.J J l !II i, r, l : LLL.) / I ,. : -.i,i. g\\ 3 l 4 ' l~~r--F4]l, l -L jdj t Ad7[ ..ii,

, p.y l i

,J., 1 4 ;, i d_;.; ' [ s t.J.f l l L SuweyUnit [ SURVEY MAP LEGEND = Survey Area Boundary - - = Suwey Unit Boundary = SuweySubunit = SurveyLocation f h g l l

4 FSV FINAL SURVEY PLAN DA T.i INTERPRETATION 4 5.0 DATA INTERPRETATION All measurements will be reported in units appropriate for comparison with the release limits. Surface activity measurements and removable contamination will be reported in units of dpm per L 2 100 cm. Gamma dose rate measurements will be reported in exposure rate units of R/hr. [. The upper boundaries of the average values for each survey unit (and/or subunit) at the 95% confidence level will be compared with the release limits. Figures 5.1 through 5.5 illustrate the logic to calculate, analyze and interpret the data for the five types of final surveys. 5.1 Conversion of Measurements to Reporting Units 5.1.1 Direct Measurements - Total Surface Activity j Measurements of total surface activity will be converted from observed gross counts per minute to net dpm/100 cm. By subtracting the background counting l 2 rate for the instrument and correcting the net count rate for geometry and 2 efficiency, the results in dpm/100 cm units are obtained. Total surface activity measurement results will be reviewed to ensure that the applied background values are appropriate. s The statistical evaluation of the resulting set of measurements is examined. When the results statistically exceed the administrative action levels, the findings will be investigated to validate the results and/or determine the need for additional survey or remediation actions. The SEG spreadsheet program, CNTTIMEC is used to establish the statistical limits based on instrument type, detector efficiency, background and countmg duration. The program provides counting time, scaler alami settings, point and MDA settings in dpm and dpm/100 cm, 2 The standard deviation may be varied to reflect confidence intervals above and below the 95 percentile. 5.1.2 Removable Contamination Measurements Measurements of removable surface activity will be converted from gross count 2 rate to units of net dpm/100 cm by subtracting the background count rate of the smear counting detector and correcting the net count rate for detector geometry and efficiency. 5.1.3 Gamma Exposure Rate Measurements The gamma exposure rate data will normally be collected using Nal(TI) or similar [ micro-R meters. These meters are easy to use, are sensitive and when used with data-logging survey meters, make recording data and entry into the database easy. Results from these meters will be compared to the response of a Reuter-Stokes pressurized ion chamber (PIC). The PIC is an industry recognized, standard 2/17/94 5-1 REVISION 0

E FSV FINAL SURVEY PLAN DATA INTERPRETATION b quality device which measures gamma exposure rate (or exposure) at environmental levels with negligible energy dependence. The PIC will be used to establish the gamma response for the micro-R meters. A correction coefficient will be calculated and applied to each micro-R meter to correct the reading to the PIC value. [ 5.1.4 Soll and Water Activity Measurements Soil and water samples will be collected in accordance with approved procedures as discussed in Section 4.3.8. Laboratory analysis will be performed to identify the presence of licensed materials. The concentrations of licensed materials in soil and water will be reported in units appropriate for comparison to the limit for annual total effective dose equivalent. 5.1.5 Ilard to Detect Nuclide Measurements Sampling of surfaces, structures and systems for HTDN will be performed as discussed in Section 4.3.10. The concentration of HTDN will be determined through laboratory analysis of these samples. If concentrations of HTDN are observed which could significantly affect previous estimates of the annual TEDE, a scenario analysis will be performed using methods and dose conversion factors provided in NUREG/CR-5512, Volume 1. 5.2 Comparison With Release Limits The method outlined below will be used to demonstrate attainment of the release limits. 5.2.1 Attainment of Release Limits for Surface Contamination

a. Total Surface Activity (fixed plus removable contamination):

Individual measurements: Not to exceed 15,000 dpm/100 cm 2 (surface area not to exceed 100 cm ), 2 Survey Unit - Random or systematic Upper limit of confidence interval sampling: (Equation 5.1 or equivalent) for the sample mean value does not exceed 5000 dpm/100 cm, 2 Survey Unit - Diased sampling: Upper limit of confidence interval (Equation 5.1 or equivalent) for the sample mean value does not exceed 2 5000 dpm/100 cm, 2/t7/94 5-2 REVISION 0

b FSV VlNAL SURVEY PLAN DATA INTERPRETATION I

b. Removable Surface Contamination:

2 Individual measurements: Not to exceed 1000 dpm/100 cm,

c. Calculation of Upper Limit of Confidence Interval of the Mean:

U, = x + t 1 (5.1) g Ei where: U, = upper confidence limit of sample mean x = sample mean value ta,, = student t statistic for the degree of confidence and degrees of freedom; df(degrees of freedom) is equal to n - 1; and a is [ 0.05 for this test S, = estimate of the sample standard deviation i n = number of measurements in the sun'ey unit 5.2.2 Attainment of Gamma Exposure Rate Limit Individual measurements: Net ganuna exposure rate does not exceed 10 pR/hr above natural background. Average measurement - buildings Gamma exposure rate does not exceed 5 pR/hr above natural background when averaged over 2 10 m, Average measurement - environs Gamma exposure rate does not exceed 5 pR/hr above natural background when averaged over 2 100 m. Individual gamma exposure rate measurements within the survey unit will be compared to the appropriate local gamma background (mean or median)= determined as described in Section 4.4.3.f. Some survey units contain construction materials and shielding components which result in a large standard ( deviation of the gamma exposure rate measurements, thus making high values [ 2/17/94 5-3 REVISION 0

FSV FINAL SURVEY PLAN DATA INTERPRETATION ( exceeding the limit probable. In order to avoid frequent investigations of probable conditions, the Student's "t" statistic is used to estimate the 99.9th percentile (one chance in 1000) for the distribution of the set of individual measurements which may then be used as a statistical test on the individual measurement. Using this test, the individual measurement must exceed the 99.9th percentile value as well as exceeding 5 pR/hr above background for the measurement to be deemed in excess of the limit. 5.2.3 Attainment of Soil and Water Limits Sampling and analysis will be performed using approved procedures as described in Section 4.3.8. The analysis results for individual samples will be compared to the background levels as determined in Section 4.4.3.e. Samples indicating the - presence of licensed materials statistically above background levels will be evaluated using the methodology contained in NUREG/CR-5512, Volume 1 to ensure that the annual TEDE will not exceed 10 mrem. Samples may exceed the 10 mrem annual TEDE limit by up to a factor of 3 providing that the average concentration within the 100 m contiguous area will result in an annual TEDE 2 that does not exceed (100/A)ir2 times the annual TEDE limit where A is the area ( of elevated activity in m. 2 5.2.4 Attainment of IITDN Limits I The ratios of IITDN to readily detected radionuclides will be compared to Previously established ratios. Where differences are observed which could r [ indicate a significant increase in contribution to the annual TEDE, scenario analyses will be performed to detemiine the added contribution to the annual TEDE and to ensure that the annual TEDE from HTDN will not exceed 2 mrem. If the calculated dose due to HTDN does not exceed 2 mrem / year, remediation solely for the reduction of HTDN will not be required. If the dose equivalent due L to IITDN is greater than 2 mrem per year, further evaluation will be performed to determine the dose equivalent for IITDN plus all contaminants. If the total (- do.se equivalent for the scenario does not exceed 10 mrem per year, no additional remediation will be required. I ( L J l^ 2/17/94 5-4 REVISION 0

YSV FINAL SURVEY PLAN DATA INTERPRETATION FIGURE 5.1 Total Sinface Contnminnfion Data Processing AfTected Stuvey Units Perform Convert data to net addtional surveys dprrY100 sq cm Remediate St:tuticany evaluate N f and conpara each value to Mon Level 4 List all measurements . above Administrative a able Action Levels U h if the value is s;100% n y of the release lirnit, the (1): > Acton Lewi value may be compared 47 directly to Level 3 n Irmsogate by management action. measurements; te y Collect addtonal SU-Surwy Unit data Compare survey unt rnean to f Acton Lewi 3 Statreca!!y enluate and calculatelocal l [. area average over Yes te 1 sq meter o V T [ Mean > Acbon r Levef 67 U ) SU pe meets retease ] critena 7 f Yes Admmistrative Action Leve!s u Acton Lewi 3. Average surface mntamination 375% of release timrt Prepara Final Acton Lewi 4. Moorriam surface contarrunaton readng >75% of release hmt Data Report ( Acton Lewl 6 - Average surface contamsnation >100% of release hrrut l [ l [ 2/17/94 5-5 REVISION 0

FSV FINAL SURVEY PLAN DA TA'INTMPRETA TION I FIGURE 5.2 Total Stufxe Contamination Data Pmcessing Unaffected StnTey Units Conwrt data to net dWOO q cm NOTES u Calculate and Ust all rneasurenwnts ahhnhh compare to Action bei2 Acton Lewls SU - Surwy Urvt u Redassfy, renwd. ate if necescary and rearvey as Affected army urut Yes Define extent of area, Estabish new subunit a Calculata summary datsbes 1f No SU meets release criteria ? 1 I Mninistrative Action Levels Prepare Final Acton Lewi 2 - Awrage arface contarranaton >50% of releam lirrut Data Report j i l I l 2/17/94 5-6 REVISION 0

FSV FINAL SURVEY PLAN DA TA INTERPRETATION-FIGURE 5.3 Removable Surface Contamination Data Pmcessing Convert data to not dpm/100 sq cm Evolume and compare each value to Acton Levet 1 Reclass#y, remedente. & resuney es ANected p su*vey un$ f SU > Acton Level 17 h Define extent of area, estatWsh new sutong es Remedete measurements or

• mvestigaton

(,) u.e -o,i o, area, estabesh new 1 subund No Affectas SU 7 Yes compare to Actm Lewl5 50 > Acton Yes = <.,7 u Yes Ac No u 50 no Criter p NOTES Prepare Final n Ust all measurements above Administrative Administrative Acton Levels SU - Survey Unit Action Level 1 - Removable contaminaten > 25% of release limit Acton Level 5 - Removable contaminaton > 100% of release Imt ( ( 2/17/94 5-7 REVISION 0

FSV FINAL SURVEY PLAN DA TA INTERPRETA TION FIGURE 5.4 ( Gamma Exposure Rate Data Pmcessing ( ( Convert data to rmcro4t/hr Calculate Perform a&Monal Wrmnd & sum SD at 99 9% CL Comsare each value @M Remects's to background & SD at 99 9% CL Ust all measurements above Administrative Ddme enent d y Vahm > Acta area. estatesh new

• 1 L/

Led 7 ? SU - Survey Unit ,umn, d No SD - Standard Deviation CL - Confidence Level f i I A Yes f values

• Acts No (1) h Le487 inwstigste measurements, Cdlect s&ttmal data CalcuWe SU mean exposure rate oar 10 sq meters i

Releas criteria Yes No Acta led B satisfied, Prepare Final ) ( 7 hubMd [ / Adrninistrattve Action Levels Acton Level 7 -Individual exposure rate >10 microR/hr Acton Level 8 - Average exposure rate >5 meroRhr r l 2/17/94 5-8 REVISION 0 r

( FSV FINAL SURVEY PLAN DATA INTERPRETATION FIGURE 5.5 ( Soil Sample Data Pmcessing mm Cda.:t sampks Populatson st$stics, UCL, { calculated only for soil sampling survey units Other soil samples Anarge san;ms evaluated individtally and reported k for endividal S.U. Saan rates, shtiy llD. FSV REMP lower limit karmed malses atne of dntaction tXrJgturd Maete > g tyrigcur:1? Rernedate Yes Calcune snud TECE fcr nutkies > Ofm estert d tursground & crsmae sea estathsh sarple to Actxn Lew g rew stburst .L TECE > Adtri Law 9 ? Yes Convrre nr:h sarnpa to 3 x Adai Low 9 TEDE > 3 x Yr5 ? un tfw 9 ? Pb i MWe 1% m p & test sq rt d (100A) j l 1r CalcuMe e,w m UCL W Aw TE M Ya for al sai > test 7 wrrem U UCL Ya rnm11 ( ( Mrdnistrative Action Levels te Action Level 9 Annual TEDE in excess of 10 mrem i gp,,pgg Data Report i 1 2/17/94 5-9 REVISION 0

VSV FINAL SURVEY PLAN FINAL SURVEY REPORTING 6.0 FINAL SURVEY REPO.RTING Upon completion of each major Final Survey phase an interim report will be prepared for review by the Nuclear Regulatory Commission. This report will meet the intent of Pegulatory Guide [. 1.86 for Final Survey reporting and will: { Identify the premises Demonstrate that reasonable efforts have been made to reduce residual contamination to ALARA levels ( Describe the scope of the Final Survey and general procedures followed j Present the results of surveys in terms of the applicable guideline values l ( Upon completion of all Final Survey phases, a Final Report will be prepared which addresses the topics outlined below. The report will provide adequate data and discussion of each topic to meet the intent of Draft NUREG/CR-5849. 6.1 Topical Outline The following outline illustrates a general format for the Final Report regarding document volumes, Wical outlinc and content. The outline below may be adjusted to provide a clear presentatitt of the infonnation. The level of detail will be sufficient to clearly describe the Final Survey Program and certify the results. The applicable nuclides will be identified. The report will demonstrate that the instrumentation selected [ was appropriate to reliably measure radioactivity at the guideline values. The report will be able to conclude that residual activity is below the applicable guideline values. ( 1.0 Background Information l 1.1 Reason for Decommissioning ( 1.2 Management Approach 2.0 Site Descriptien 2.1 Type and Location of Facility 2.2 Ownership f 2.3 Facility Description Facility Grounds Facility Structures [ Facility Systems l r 3.0 Operating IIistory i l 3.3 Licensing and Operation 3.J Processes Performed [- 3.3 Waste Disposal IIistory and Practices ( 2/17/94 6-1 REVISION 0

I

• ~

VSV FINAL SURVEY PLAN FINAL SURVEY REPORTING 4.0 Decommissioning Activities s 4.1 Objectives 4.2 Results of Previous Surveys Site Characterization ( Radiological Environmental Monitoring L~ Radiological Effluent Reports r 4.3 Decontamination and Dismantlement Activities L 4.4 Decommissioning Processes Decommissioning Contractor r Demolition and Dismantling L Shipping, Storage and Disposal of Materials Security Precautions and Safeguards f 5.0 Final Survey Methodology L 5.1 Sampling Parameters 5.2 Background Levels 5.3 Major Contaminants Identified y. L 5.4 Guidelines EstaF"shed 5.5 Equipment and Techniques Applied Survey Instruments and Equipment c L Detection Sensitivity 5.6 Survey Process Survey Unit Classification Reference locators Surface Scans Surface Activity Measurements Exposure Rate Measurements Soil, Water and Sediment Sampling Special Sampling ] 5.7 Survey Control Quality Control and Quality Assurance J Personnel Training and Qualification Instrument Calibrations Custody of Data Data Verification Survey Records 5.8 Data Analysis and Statistical Evaluation ? )

ze,,4 e.2 REvISle s.

FSV VINAL SURVEY PLAN Fl? J. SURVEY REFORTING ~ 6.0 Final Survey Results 6.1 Data Results and Interpretation Data Reduction and Review Statistical Evaluation 1 6.2 Comparisons to Release Criteria 6.3 Comparisons to Independent Third-Party Survey 7.0 Summary APPENDIX I - Survey Unit Release Records 1. Replicate Surveys 2. Survey Maps APPENDIX II - Westinghouse Team Quality Assurance Documentation 1. Qualification Records 2. Quality Assurance Surveillances [ 6.2 Reporting of Survey Findings f 6.2.1 Summary l l Measurement results will be reported at several levels of detail. A summary of I the measurement results and overall conclusions showing that the facility meets j the release criteria will be provided. As applicable, a tabular data summary will present the results for each major category of survey unit such as: stmetures, outdoor areas and plant systems. This tabulation will identify the number of survey units, the number and type of measurements such as: total surface beta-gamma, total surface alpha, removable surface beta-gamma and removable surface alpha activity concentration, and gamma exposure rate. Average and 2 maximum values of surface contamination in units of dpm/100 cm and upper limits of confidence intervals about the mean are reported for comparison to the release criteria surface activity limits in Table 3.1. A summary of gamma j exposure rate measurements will be presented by similar ticatment. These results { will likely be included as graphs that will better illustrate the individual data j points and the statistical distribution of the results. t l 1 6.2.2 Summary Data Reporting for Each Survey Unit r - Within the release record for each survey unit (and/or subunit), the number of L measurements and the applicable statistical distribution will be preeented in graph 2 form. These will be reported in units of dpm/100 cm for each type of measurement: total surface beta-gamma, total surface alpha, removable surface [E ' eta-gamma and removable surface alpha activity concentration. For each survey -2/17/94 6-3 REVISION 0

FSV FINAL SURVEY PLAN FINAL SURVEY REPORTING unit (and/or subunit) gamma exposure rate measurement results will be reported showing the number of measurements and the appropriate statistical attributes (- relative to each survey unit (and/or subunit). The use of applicable gmpisics will illustrate the required statistical evaluation and acceptance criteria. The applicable results of routine and special sampling measurements, e.g., sediment, paint, concrete and other debris will be reported in the release record for each survey unit. 6.2.3 Detailed Data Reporting The result of each measurement taken in the Final Survey will be graphically presented. Graphs and tables of similar format will be used to report data for ( systems and outdoor survey units. Final data reports provide indication of individual measurements which exceed the guidelines. The release records for survey units classified as " unaffected" will include individual measurements which (~ exceeded the predetermined administrative action level and the results following the investigation of these measurements. f I ( ( (- [ [ -2/17/94 6-4 REVISION 0 _ _-___-_ - -____ ____ ______ ____=

b 4 FSV FINAL SURVEY PLAN REFERENCES

7.0 REFERENCES

1. NUREG-0586, " Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," August 1988. 2. "Fon St. Vrain Decommissioning Plan", Public Service of Colorado, November 05, 1990; Revised July,1991; Revised April,1992; Approved November,1992. 3. Annual Radiological Environmental Report, Summary Report for the Period January 01, 1991 - December 31,1991", Public Service of Colorado, April 24, 1992. 4. " Annual Radiological Environmental Report, Summary Report for the Period January 01, 1990 - December 31,1990", Public Service of Colorado, April 12, 1991. 5. " Fort St. Vrain Initial Ra<liological Site Characterization Report", April 30, 1992. 6. "Results of ISFS1 Site Background Radiation Study", Department of Radiology and Radiation Biology, Colorado State University, November 02,1990. 7. U.S. Atomic Energy Commission, Regulatory Guide 1.86, " Termination of Operating Licenses for Nuclear Reactors," June 1974. 8. " Supplement to Applicant's Environmental Report Post Operating License for Proposed Decommissioning of the Fon St. Vrain Nuclear Station", April,1992. 9. J. Berger, NUREG/CR-5849, " Manual for Conducting Radiological Surveys in Support of License Termination," June 1992 (Draft). 10. PSC Letter, Warembourg to Austin dated December 23,1993,

Subject:

" Final Radiation Survey Plan, Treatment of Hard to Detect Nuclides."

11. NRC Memorandum P.B. Erickson (NRC) to Seymour H. Weiss (NRC);

Subject:

" Summary of Meeting with Public Service Company of Colorado (PSC) To Discuss Preliminary Decommissioning Plan of June 30, 1989," dated August 24,1989. ( 12. NUREG/CR-5512, U.S. Nealear Regulatory Commission, " Residual Radioactive Contamination From Decommissioning", Volume 1, October,1992. (. 13. Fort St. Vrain Decommissioning Project Radiation Protection Manual, Scientific Ecology Group. c 1 ( -- 14. J P. Andrews, " Count Time Spreadsheet, SEG CNTTIME@", Scientific Ecology Group, } Inc., Revision: August 8,1993. 2/17/94 - 7-1 REVISION 0

FSV FINAL SURVEY PLAN REFERENCES 15. Fort St. Vrain Nuclear Station Radiological Environmental Monitoring Program Annual Summary Report for 1989. 16. "PSC Site Characterization Data," Department of Radiology and Radiation Biology, Colorado State University,1990. [.:. 17. "Shoreham Decommissioning Project Termination Survey Plan", Shoreham Nuclear Power Station, April,1993. ( ( (- L L [. 2/17/94 7-2 REVISION 0

[ FSV FINAL SURVEY PLAN GLOSSARY 8.0 GLOSSARY Administrative Action Level-A contamination level used as the investigation threshold in survey units to evaluate the need for additional investigation, reclassification or remediation. Affected_brea - Areas that have potential radioactive contamination (based on plant operating history) or known radioactive contamination (based on past or preliminary radiological surveillance). 11ackground Radiation - Naturally occurring radiation in the human environment. It includes } cosmic rays, radiation from the naturally radioactive elements and man-made radiation from j global fallout. Ilackground Value - A numerical value, statistically evaluated and expressed in the appropriate unit of measurement, defining the background radiation. Iliased Sample - A method of selecting sampling locations which incorporates a non-random error, i.e., a method which selectively chooses locations for sampling which have a higher probability of contamination than those locations not selected. 131ngd Survey - A method of selecting survey measurement locations which incorporates a non-random error, i.e., a method which selectively chooses locations for survey measurements which have a higher probability of contamination than those locations not selected. Characterization Survey - A radiological survey and supporting evaluations performed to establish the FSV baseline radiological condition for planning decommissioning activities. The Characterization Survey activities are described in and controlled by the FSV Decommissioning Project Radiation Protection Manual. The Characterization Survey results are contained in the FSV Radiological Site Characterization Report. Component - An individual equipment item, e.g., a valve, pump, tank, or motor. _C_ontfidence Intervit!- A range of values derived from a sample such that there is a probability a, that a population parameter being estimated, e.g., a mean value, lies within the range (Ref.15). Confidence Level - The probability a, associated with a confidence interval which expresses the probability that the confidence interval contains the population parameter value being estimated. DECON - The decommissioning alternative which involves prompt removal of radioactive materials to achieve residual contamination and radiation levels which are below limits ( established to permit the facility to be released for unrestricted use. Direct Measurement - A radiological survey measurement performed by holding a detector on or close to the surface and recording the response. l /t7/94 8-1 REVISION 0

4 FSV FINAL SURVEY PLAN GLOSSARY Elevated Area Guideline - The value which individual measurements may not exceed under any conditions. Final Survey - Radiological measurements, evaluations and supporting activities undertaken to demonstrate that the FSV facility satisfies the criteria for unrestricted use. Also referred to as ( Termination Survey. Final Survey Groun-Westinghouse personnel selected to design and implement the final survey. [ Final Survey Rgno_n - A report describing the methods and results of the Final Survey. It [ initiates the NRC review and final inspection of the facility for termination of the facility license. L It is also called the Final Report. Fixed Point Measurement - A synonym for direct surface contamination measurement. Ilard-to-Detect Nuclide - (IITDN), A nuclide emitting radiation (s) of low energy or intensity such that detection utilizing typical Geld instrumentation is difficult. Also: DifGcult-to-Detect, DifGcult-to. Measure. liistory File - A compilation of information prepared for use in planning the termination survey } of a survey unit. It summarizes the operational history, characterization survey data, operational J surveys and other information to help establish the basis for the design of the Final Survey. Interim Report - A report prepared at the conclusion of a major phase of the Final Survey. The interim report will contain a description of the areas surveyed and survey results. Survey results in this report will be compiled and submitted in the Final Report. NRC - U.S. Nuclear Regulatory Commission Non-Suspect Affected Area - Affected areas are designated as non-suspect if the average contamination levels do not exceed 50% of the release criteria for total surface area contamination and, do not exceed 25% of release criteria for removable contamination. Operational Radiolonical Survey - A radiological survey performed in accordance with procedures contained in the FSV Decommissioning Project Radiation Protection Manual. Operational surveys are distinct from, and usually performed prior to, final surveys. Outdoor Area - A category of survey units which includes site grounds, outside surfaces of buildings and small structures located out of doors. Plant Structures - All FSV Nuclear Generating Station site buildings and their surfaces (generally identified as civil structures). For purposes of the Final Survey, all structures such as platforms, { f restraints, supports and other physical items are considered to be components. External surfaces i of piping systems, heating and ventilation systems, tanks, stacks, etc., are also treated as components in the Final Survey. 2/17/94 8-2 REVISION 0

FSV FINAL SURVEY PLAN GLOSSARY Por>ulatioE - A collection of all possible values of a radiological parameter being measured in the Final Survey, A survey unit (or subunit) is considered to be a population for purposes of f: drawing inferences regarding the value of a parameter, such as contamination level mean value in the entire survey unit (or subunit), based upon a sample of measured values. [ Process System Index - A listing, controlled by a station procedure, which identifies and assigns a unique identification code to each plant system. Random Senpig - In survey design, a method for selection of measurement locations whereby each of the individual locations defined in the sample area has an equal probability of being selected. Melated terms are: random selection and randomly selected. J1elease Criteriq - A term used to identify the radiological requirements for release of the FSV facility for unrestricted use. These requirements, which consist of specified limits for residual contamination and radiation levels, are specified in the FSV Decommissioning Plan. ( Itelease Limit - The principal numerical limits in the facility release criteria presented in Table 3.1. ILelease Record - A document compiled for each survey unit (stmeture, system or outdoor area) which demonstrates that it is suitable for unrestricted use. It contains evaluated survey data and supporting information to provide a concise record of the results and basis for the conclusion that the release criteria are satisfied. Reportine Units - The units in which each type of survey measurement is expressed for comparison to release criteria limits. For surface contamination measurements the reporting 2 units are dpm/100 cm and for gamma exposure rate measurements the units are R/hr. Jtestricled_2\\rea - The group of major buildings on the FSV site located inside the security fence. This group includes of the Reactor, Turbine, Radwaste Compactor, and Maintenance buildings. Sean Survey - A qualitative radiological monitoring technique which is performed by moving a detector over a surface (typically within one centimeter of the surface) at a specified constant speed to detect elevated contamination or radiation levels. Under the correct circumstances and controls a scan can be quantitative. Similar terms applied to this technique are: Scan and Surface Scan. Site Characterization Report - A report (including addenda) which documents the surveys, calculations and evaluations and presents the results of the Fort St. Vrain Initial Radiological Site Characterization. Subunil - A subunit, as used in survey design, is a subdivision of a complex surve', unit that incorporates a structure, an item of equipment, or some other feature in order to esiablish that an appropriate number of survey measurements be made within the subunit, as well as within the survey unit. 2/17/94 8-3 REVISION 0

[ FSV FINAL SURVEY PLAN GLOSSARY Survey Area - Categories or types of survey unit groupings having similar physical characteristics. Three categories of survey areas have been established: plant systems, ( structures and outdoor areas. Sutvpv Desien - The process of determining the type, location, number and frequency (or u f density) of radiological measurements to be taken in the Final Survey. p Survey Desien Guidelines - Criteria established to provide the appropriate level of survey 1 intensity for systems, structures and outdoor areas, based upon their classification. ( Survey Instructions - Written directions which specify the type and number of measurements to l be taken in a survey unit. The survey instmetions are in a standard format on forms controlled by a Final Survey procedure. Each survey package includes survey instructions. Survey Location - A discrete area or subdivision of a survey unit that is smaller than a subunit but larger than a survey point. In survey design, a survey unit (or subunit) is divided into a collection of survey locations. Specific locations are selected in accordance with the design guidelines based upon the type and classification of the survey unit. In a structural or outdoor survey unit, a location is usually represented by a single grid block. In a system survey unit, ( a specified length of piping or a component such as a valve is referred to as a survey location. A survey location can contain one or more survey points. [ Survey Package - A collection of information in a standardized format for controlling and documenting field measurements taken for the Final Survey. A survey package is prepared for each Survey Unit. The survey package includes the survey instmetions, a control form, grid ( map (s), survey measurement data sheets and survey maps. Survey Point - A smaller subdivision within an area designated as a survey location (grid block, system component) where local measurements are taken, generally referring to an area covered i 2 by a detector, or an area of 100 cm when a smear is taken. I Survey Qnit - A division of the facility consisting of a grouping of contiguous (usually) structural areas, outdoor areas, or functionally contiguous equipment items. The Survey Unit is the basic { entity for management of the Final Survey. Survey Unit Gassification Description - A listing of all survey units established for the Final Survey which identifies the classification of each as "affected" or " unaffected". Suspect Affected Area - Affected areas are designated as suspect if the average contamination levels are above 50% of the release criteria for total surface area contamination, or above 25% of release criteria for removable contrunination. i f Systematic Samole - A sample which is obtained by some systematic method as opposed to a random sample; for example, selection from a list using a specified interval for selection. In a structural survey unit which has been uniformly gridded, a systematic sample could, for example, be comprised of every fourth block. 2/17/94 8-4 REVISION 0 e

FSV FINAL SURVEY PLAN GLOSSARY o Igtal Effective Dose Equivalent (TEDID - The sum of the deep dose equivalent (for external exposui md the committed effective dose equivalent (for internal exposures). IJnaffected Alta - All areas not classified as affected. These areas are not expected to contain residual radioactivity, based on a knowledge of site history and previous survey information. [ Verification Survey - A radiological survey which consists of repeat measurements at a specified fraction of the survey measurement locations in a survey unit, usually selected at random, to [ provide an independent check of final survey measurements. Also called a replicate survey. WorUngatqtion - A document used to guide performance of a task. ( [ ( ( ( ( ( [ { 1 UlW94 8-5 REVISION 0 i .}}