Appendix — Vermont Yankee Nuclear Power Corp. v. Natural Resources Defense Council, Inc.

Supreme Court brief1978

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APPENDIX

VOLUME Il - RODAR, JR, CLERK |

Paces 401 - 1396

In THE

Supreme Court of the Anited States

Octrosper Term, 1976

No. 76-419

VERMONT YANKEE NUCLEAR POWER

CORPORATION,

Petitioner

NATURAL RESOURCES DEFENSE COUNCIL,

INC., ET AL.,

Respondents

No. 76-528

CONSUMERS POWER COMPANY,

Petitioner

NELSON AESCHLIMAN, et Au.,

Respondents

ON WRITS OF CERTIORARI TO THE UNITED STATES COURT OF

APPEALS FOR THE DISTRICT OF COLUMBIA CIRCUIT

PETITIONS FOR WRITS OF CERTIORARI FILED

SEPTEMBER 21 AND OCTOBER 14, 1976

CERTIORARI GRANTED FEBRUARY 22, 1977

401

[SUMMARY OF APPLICATION

VERMONT YANKEE NUCLEAR POWER STATION

Ex. 3, pp. 2-5, AEC Dkt. No. 50-271]

SITE AND ENVIRONS

The site for the Vermont Yankee Nuclear Power Station

is located in the town of Vernon, Vermont in Windham

County on the west shore of the Connecticut River upstream

of the Vernon Hydroelectric Station. The site consists of

about 125 acres owned by the Vermont Yankee Nuclear

Power Corporation and a narrow strip of land owned by

New England Power Company and located between the

Connecticut River and the boundary of the Vermont Yankee

property and in which Vermont Yankee has perpetual

rights and easements granted by the New England Power

Company with the approval of the Federal Power Commis-

sion. The City of Brattleboro, Vermont is about five miles

upstream of the site. There are no residences on the site,

and the minimum distance from the reactor building to the

boundary of the exclusion zone, as defined in 10 CFR 100, is

910 feet. The nearest town with a population of 25,000 or

greater is Northampton, Massachusetts, about 30 miles to

the south. Much of the land around the site is undeveloped.

Most of the developed land is used for agriculture and dairy-

ing, with houses scattered or grouped in small villages. Ap-

pendix B appended hereto is a map showing the location of

the site with respect to adjacent areas (FSAR 2.2).

In July 1969 Vermont Yankee initiated a comprehensive

environmental radiation surveillance program designed to

monitor the atmospheric, terrestrial and aquatic environ-

ments in the vicinity of the station. This program was de-

veloped in cooperation with the Vermont State Department

of Health and reviewed by the Vermont Water Resources

Board.

402

It is Vermont Yankee’s intent that the environmental sur-

veillance program will be closely coordinated with condi-

tions of station operation by systematically correlating in-

formation developed through continuous in-plant monitor-

ing, effluent release records and environmental sample col-

lection and analysis. The primary purpose of the surveil-

lance program is to make station management and the State

Health Department aware of any conditions of station op-

eration which may have a significant effect on the environ-

ment. The operational surveillance program will be a

graded program subject to periodic critical review (FSAR

2.6).

Plant Design Bases Dependent Upon Plant Site

and Environs Characteristics

Information relating to the site and environs for the

Vermont Yankee Nuclear Power Station is presented in

Section 2 of the FSAR. The design features which were

dependent or affected by the site characteristics are sum-

_ marized below.

Condenser Off-Gas System

The condenser off-gases will be discharged to the atmos-

phere through deeay piping and a stack 308 feet above

grade. The off-gas system is used during normal operation.

It is expected that the release of gaseous radioactivity will

be kept within small fractions of the limits specified in 10

CFR 20 based on the use of the off-gas system. However,

to further minimize the release of gaseous radioactive

wastes, Vermont Yankee will modify the off-gas system to

provide additional holdup and subsequent decay of air

ejector off-gases. The design of this advanced off-gas sys-

tem is currently underway and fabrication and installation

403

will be completed by the end of the first refueling outage

(FSAR 9.4 and Appendix I).

Liquid Waste Effluents

Liquid effluents will be released from the radwaste dilu-

tion system at a point downstream of the aeration blocks in

the station discharge structure. The liquid waste treatment

system has been designed so that those liquid effluents

which are released from the station are always well within

the limits specified in 10 CFR 20. This is a batech-type sys-

tem wherein the wastes are separately collected and pro-

cessed based upon the most efficient methods, Batches of

liquid radwaste are sampled and analyzed prior to dis-

charge. If suitable for discharge, radwaste effluents are

released at a rate such that the resultant concentration at

the point of release will never exceed 10 CFR 20 nor the

applicable state release limit. During each release, a con-

tinuous radiation monitoring device is set to alarm upon

excessive activity concentration in the radwaste discharge

line. Diseharge records are maintained for all station

liquid radwaste effluents.

Wind Loading Design

Plant structures are designed to withstand the effects of

80 mph winds with gusts to 110 mph. Although the prob-

ability of a tornado occurring at the Vermont Yankee site

is small, all structures and equipment necessary to initiate

and maintain a safe plant shutdown are designed to with-

stand a tornado wind loading of 300 mph (FSAR 2.3 and

12.2

IHlydrology

The site is at mile 138.3 above the mouth of the Connecti-

eut River on the west bank on the pond formed by the

404

Vernon Dam and Ilydroelectriec Station. The maximum

recorded water level at Vernon was 231.4 feet mean sea

level which oceurred during the flood of March 19, 1936.

The minimum elevation of access openings to buildings

housing important equipment is. 252.5 feet mean sea level,

more than 20 feet above the maximum observed water level.

An analysis of water levels resulting from the probable

maximum flood, a postulated occurrence which is more

severe than the most significant historical event, has been

performed and shows that the station is suitably protected

against the maximum probable flood (FSAR 2.4 and Ap-

pendix IL).

Geology

The site area is overlaid by glacial deposits, with an

average of 30 feet of glacial overburden above the local

bedrock. Major structures of the station, including the

reactor building and turbine building, are supported on

rock. Seismic velocity measurements at the site verified

the hard massive nature of the bedrock and there were no

indications of deeply weathered or faulted zones (FSAR

2.5).

Seismic Design

The maximum acceleration at the bedrock surface of the

plant site from an earthquake is expected to be from 0.03g

to 0.04g. Although this acceleration is considered to be con-

servative, the seismic design for structures and equipment

important to safety are based on dynamic analyses using

acceleration response spectrum curves together with a

ground motion of 0.07g. The station is also designed so that

the plant can be shutdown in a safe and orderly manner

even if the ground acceleration were 0.14g (FSAR 2.5 and

12.2).

405

SUMMARY DESCRIPTION OF THE FACILITY

3.1 Lwrropuction

The nuclear steam supply system of the Vermont Yankee

Nuclear Power Station utilizes a General Electric boiling

water reactor identical in most design features to Monti-

cello Unit No. 1, the operation of which was recently author-

ized by the Atomie Energy Commission, The design of the

features of the Vermont Yankee facility is based upon

power technology attained during the development, design,

construction and operation of boiling water reactors of

similar types, including the Monticello facility. These fea-

tures include the basic fuel design, Zircaloy fuel cladding,

hydraulically operated control rods, in-core neutron moni-

toring instrumentation, pressure suppression containment

and radioactive waste control.

The following discussion summarizes the principal de-

sign features of the Vermont Yankee facility which are sig-

nificant to safety considerations. The FSAR and subse-

quent amendments present an extensive technical deserip-

tion and evaluation of this facility and FSAR references

elaborating on the discussion below are given in each sum-

mary. Appendix C of this testimony gives a comparison of

the significant station design features of Vermont Yankee

to other boiling water power reactor stations. A drawing

of a reactor and containment typical of the Vermont Yankee

plant is shown in Appendix D.

3.2 Reactor Primary System

The reactor is a single eyele, forced circulation, boiling

water reactor producing steam for direct use in the steam

turbine.

The fuel for the reactor core consists of slightly enriched

uranium dioxide pellets contained in sealed Zircaloy-2

406

tubes. These fuel rods are assembled into individual fuel

assemblies of 49 fuel rods each, Each complete core load-

ing consists of 368 fuel assemblies (FSAR 3.2),

Control of the reactor is achieved by movable control

rods. Reactor power level control is augmented by con-

trolling the recirculation flow rate through the reactor core.

The core is designed so that the reactor can be shutdown

from any point in its operating cycle by control rods alone,

Stored energy will be available from reactor pressure and

gas-charged accumulators to insert all control rods simul-

taneously for rapid shutdown of the reactor (FSAR 3.4).

In addition, there is provided a standby liquid control

system containing a boron neutron-absorbing solution which

is capable of shutting down the reactor and maintaining it

in a shutdown condition. This system is an independent

system that would be used to shutdown the reactor in the

unlikely event that shutdown cannot be accomplished with

the control rod system alone (FSAR 3.8).

The reactor pressure vessel contains the reactor core and

supporting structure, the steam separators and dryers, the

jet pumps, the control rod guide tubes, the feedwater and

core spray spargers and other components. The main con-

nections to the reactor vessel include the steam lines, reactor

coolant recirculation lines, feedwater lines, control rod drive

housings and other connections fortore cooling. Appendix

I. is a cut-away view typical of the reactor vessel and core

arrangement (FSAR 3.3).

Reactor Coolant will enter the reactor near the bottom of

the core and flow upward through the fuel assemblies. The

heat from the nuclear reaction will produce a mixture of

steam and water. The steam will be separated from the

water by means of steam separators and dryers located in

the upper portion of the reactor vessel, and will then pass

407

through steam lines to the turbine. The separated water

will mix with the incoming feedwater within the reactor

vessel and the total flow will be returned to the core inlet

through jet pumps located in the vessel. The motive force

for the jet pumps will be supplied by the water discharged

from two reactor coolant recirculation pumps. Variable

frequeney motor-generator sets supply power to the recir-

culation pump motors.

The design, fabrication, erection, testing, and certifica-

tion of the reactor pressure vessel are discussed in detail

in Amendment No. 13 to the license application (applica-

tion dated December 2, 1966, Docket No, 50-271).

The plant has removable insulation in specified areas

which permit direct inservice inspection of the exterior of

the reactor vessel and other primary system components.

This permits inspection of selected nozzle-to-shell welds,

inspection of reactor vessel studs, nuts and bushings, se-

lected sampling of high eyelie stress vessel welds. and

selected sampling of coolant system piping, pumps, valve

supports, and hangers. Inspection of the interior of the re-

actor vessel can be accomplished under water using re-

motely operable lights and visual aids such as boresecopes

and underwater TV cameras. (FSAR 4.2.6, 4.3.6, 4.4.8,

and Section 4.6 of the Vermont Yankee Technical Speei-

fications. )

3.3 Plant Containment System

The primary containment system, consisting of a steel

light-bulb-shaped drywell, a steel doughnut-shaped pressure

suppression chamber, and interconnecting vent pipes, pro-

vides the first containment barrier surrounding the reactor

vessel and reactor primary system. Any leakage from the

primary containment system is to the secondary contain-

408

ment system which consists of the reactor building, the plant

standby gas treatment system, and the plant main stack.

The integrated plant containment system and its associated

engineered safeguards features is designed so that off-site

doses resulting from postulated design basis accidents are

well below the reference values stated in 10 CFR 100,

3.3.1 Primary Containment System

The primary containment is designed to accomodate the

pressures and temperatures which would result from, or

occur subsequent, to, a failure equivalent to a double-ended,

circunferential rupture of a reactor coolant recirculation

system line within the primary containment resulting in

the loss of reactor water at the maximum rate. The pres-

sure suppression chamber is a steel, torus-shaped pressure

vessel approximately half filled with water, and... .

[USAEC STAFF SAFETY EVALUATION,

VERMONT YANKEE NUCLEAR POWER STATION,

Ex. 9, p. 5, AEC Dkt. No. 50-271)

. connected by vent pipes to the doughnut-shaped

steel water-filled suppression chamber (torus). The reactor

vessel and the recirculation water piping, valves and pumps

are housed in the drywell. The reactor is a direct-cycle,

foreed circulation, boiling water system which produces

steam at 1000 psig. The reactor core uses slightly enriched

uranium dioxide fuel in the form of pellets contained in

Zirealoy tubes. Water, which serves as both the moderator

and coolant, is pumped upward through the core where it

boils due to the heat produced in the core. Two recircula-

tion loops, each with a pump, provide the driving flow. The

resulting steam is supplied to the turbine, which is directly

coupled to the generator. After passing through the tur-

bine, the steam is condensed, partially reheated and re-

turned to the reactor.

409

SITE AND ENVIRONMENT

Site Description

The plant is located near the village of Vernon, Vermont,

slightiy less than four miles north of the Massachusetts

state line. It is on the west bank of the Vernon Pond. The

pond is formed behind the Vernon Hydroelectric Station

dam one-half mile downstream on the Connecticut River.

The west bank of the river forms the boundary between the

states of Vermont and New Hampshire flows in a south-

easterly direction in a narrow, steep-sided valley through

the Pisgah Mountains. The terrain rises steeply from the

valley floor in the vicinity of the plant and reaches the

same elevation as the plant stack at a point 6300 feet from

the plant. The site is bounded on the west by the village

and on the....

410

UNITED STATES OF AMERICA

ATOMIC ENERGY COMMISSION

IN THE MATTER OF

VERMONT YANKEE NUCLEAR Docket

POWER CORPORATION No. 50-271

(VERMONT YANKEE NucLEAR Power STaATIon)

ORDER DETERMINING OBJECTIONS

TO INTERROGATORIES

On April 30, 1971, Intervenor Natural Resources De-

fense Council (NRDC) submitted various interrogatories

to Vermont Yankee Nuclear Power Corporation, the Ap-

plicant herein. Objections to these interrogatories were

filed by the Applicant which emphasized the difference

between interrogatories and motions for production of

documents, as defined in the Federal Rules of Civil Pro-

cedure and as adapted by the Atomic Energy Commission

to its Rules for proceedings. Applicant, after stating its

general objection based upon this difference, then waived a

portion of its objection by agreeing to make requested docu-

mentary material available for inspection and copying by

the Intervenors.

The Atomie Safety and Licensing Board adheres to its

ruling respecting the general objection made by Applicant

to interrogatories submitted by New England Coalition on

Nuclear Pollution in its Order dated May 11, 1971, and

reasons therefor will not be repeated but are incorporated

as fully as though set forth herein.

411

Applicant, in addition to a general objection which is

sustained tothe extent of the Order respecting New Eng-

land Coalition on Nuclear Pollution interrogatories, also

made specific objections to two main categories of inter-

rogatories by NRDC which, however, were limited to one

subject respecting the transportation and disposal of high

level nuclear waste materials which will oceur after leay-

ing the plant site. The objection was upon the ground that

off-site activities respecting such waste materials were not

within the issues prescribed by the Commission for this

proceeding.

Upon a consideration of the interrogatories and objec-

tions thereto,

IT IS ORDERED, in accordance with the Atomic Energy

Act, as amended, and the Rules of Practice of the Commis-

sion, that the specific objection made by the Applicant

respecting the transportation and disposal of high level

waste products is sustained upon the ground that such

considerations are not within the issues prescribed by the

Commission for this proceeding, but rather are the sub-

jects for other licensing activities.

ATOMIC SAFETY AND LICENSING BOARD

By Samuei W. Jenscu, Chairman

Issued :

May 11, 1971

Germantown, Maryland

AEC Docket No. 50-271

APPLICANT’s ANSWERS TO

INTERROGATORIES OF

NaTuRAL Resources Derense CouNciL

SERVED ON

Vermont YANKEE NuCLEAR Power CorPoRATION

413

UNITED STATES OF AMERICA

ATOMIC ENERGY COMMISSION

IN THE MATTER OF

VERMONT YANKEE NUCLEAR Docket

POWER CORPORATION No, 50-271

(VERMONT YANKEE NUCLEAR POWER STATION )

AFFIDAVIT OF DONALD E. VANDENBURGH

Now comes Donald E. Vandenburgh, and being on oath,

deposes and says as follows:

That he is a Vice President of Vermont Yankee Nu-

clear Power Corporation.

That attached hereto, are the answers to the inter-

rogatories propounded in the above entitled matter

to Vermont Yankee Nuclear Power Corporation by

Natural Resources Defense Council; and that there

is also attached certain “Supplementary Information

—IV Transportation: High Level Wastes from the

Plant” which should be considered as supplementing

the answers given with respect to Interrogatories

Nos. [V.1-6, [V.(DIS).2, 46.

That in the preparation of these answers, he has con-

sulted with the following individuals with regard to

the questions set forth next to their names below:

Name Numbers

R. M. Grube 1.1-7, V.1-18

R. J. Merlino IT.1-3, TV.1-22

IV(DIS) 1-8

W. P. Johnson IIT.1-9

J. W. Stacey IV(DIS) 9 and 10

414

4. That the answers attached hereto, are true and cor-

rect to the best of his knowledge and belief.

5. That on the advice of council and in view of the rul-

ings of the Board with respect to certain objections

sustained to the interrogatories, no answers are in-

cluded for questions 1.7 and IV.(DIS) 8.

DonaLp E. VanpENBURGH

DonaLtp FE. VanpENBURGH

COMMONWEALTH OF MASSACHUSETTS

Worcester, §.S. June 1, 1971

Then appeared before me, the above subscribed Donald

Kk. Vandenburgh and made oath that he is a Vice President

of Vermont Yankee Nuclear Power Corporation; that he is

authorized to subscribe to the answers of that corporation

to the interrogatories of the Natural Resources Defense

Council; that he has read the foregoing affidavit; and that

the statements set forth therein are true to the best of his

knowledge and belief.

Before me,

ArManpd R. Soucy

Armand R. Soucy — Notary Public

My Commission Expires September 9, 1977.

415

I. BASIC INFORMATION

QUESTION 1.1

How long is it expected that the Vermont Yankee plant

will run before being shut down for fuel reloading and

removal of high level wastes from the reactor core?

ANSWER IL.1

The first evele of plant operation is expected to run for

about 18 months before shutdown for refueling. At the

first shut down, about 4% of the fuel assemblies will be

removed and stored for reinsertion in a later cycle. At the

end of the second cycle of about 12 months, about 409% of

the fuel assemblies will be removed for shipment and re-

processing.

QUESTION 1.2

How often is it expected that refueling and removal of

high level wastes will be necessary?

ANSWER 1.2

For cycles subsequent to the second cycle, it is expected

that shutdowns for refueling and removal of spent fuel

assemblies will occur about every 12 months when about

25-30% of the assemblies will be removed.

QUESTION 1.3

Where will the high level wastes removed from the

reactor core be stored prior to their removal from the plant

site?

ANSWER 1.3

After removal from the reactor core, irradiated fuel as-

semblies will be stored in the spent fuel storage pool under

23 feet of water shielding. Spent fuel storage is described

in Section 10.3 of the VY FSAR.

416

QUESTION L4

llow long will the wastes be stored at the site before

they are removed?

ANSWER L4

Spent fuel assemblies are normally stored for a minimum

of 3 to 4 months prior to shipment to a reprocessing plant.

It is not expected they would be stored longer than a year

prior to shipment.

QUESTION 1.5

By what means of transportation will the wastes be taken

away from the plant? Lf this is expected to change during

the lifetime of the plant, describe each means and the period

over which it is expected to be used.

ANSWERS 1.5

The means of spent fuel transportation has not been

selected. VY facilities can accommodate both truck and

rail transportation. Any means of transportation will be

in compliance with government regulations. During the

lifetime of the plant, more than one means of transporta-

tion could be used.

QUESTION 16

lf the wastes are to be reprocessed, at what plant or

plants will this be done? If this is to change over the life

of the plant, specify the periods during which each re-

processing plant or plants is/are to be used.

ANSWER 16

It is planned that all VY spent fuel will be reprocessed.

A plant for reprocessing VY’s spent fuel assemblies has

not been selected. The following reprocessing plants are

now in operation, under construction, or planned for the

417

time period when the first batch of VY spent fuel will need

to be reprocessed.

1. The Nuclear Fuel Services, Inc. plant in operation,

at West Valley, New York.

2. The General Electric Co. plant under construction, at

Morris, Illinois.

3. The Allied-Gulf Nuclear Services plant, under con-

struction, at Barnwell, S.C.

4. The NUMEC Atlantic-Richfield plant, planned to be

located at Leeds, S.C,

Over the life of the VY plant, several different reprocess-

ing plants could be used.

QUESTION 1.7

If the wastes are to be reprocessed at a reprocessing

plant :

a.) Deseribe where the resulting reprocessed waste ma-

terials will be taken from the reprocessing plant. [n-

elude deseription of where reprocessed materials to be

disposed of are to be taken, and where any other

materials are to be taken. If this is expeeted to

change during the life of the plant, give this deserip-

tion for each relevant period.

b.) By what means of transportation will each segment

of the reprocessing wastes be taken to each of these

destinations? If this is to change over the period of

the plant’s life, specify the means during each phase

of the plant’s life.

ce.) Where will the waste products which must be disposed

of be taken? If this is to change during the life of the

plant, describe where the wastes are to be taken dur-

ing each phase of the plant’s life.

418

ANSWER 1.7

We are advised by legal council that in view of the

Board’s findings, no answer is required.

QUESTION. I1.1

Describe in detail the procedures to be used to remove

the radioactive materials from the reactor in order to

allow for refueling, including, but not limited to:

a.) How is the reactor core vessel to be opened and

processed to assure that no radioactivity escapes

from the containment vessel during this process?

b.) What is the chemical and radiological composition of

the materials which will be removed from the core at

that time?

ce.) What toxie chemical compounds will be present in this

process, in what quantities and proportions?

d.) What radioisotopes will be present in the material

removed from the reactor core at this time? List them,

including the quantity expected of each, the propor-

tion of the isotope to the total radioactive isotopes

present, the total amount of radioactivity represented

by each type of isotope (expressed in appropriate

nits), and the half-lives of each isotope.

ANSWER IL1

a.) The reactor is first shut down and cooled down. The

top of the primary containment vessel is removed to

expose the top of the reactor vessel. The bolted head

of the reactor vessel is removed and the area flooded

with about thirty feet of water. All subsequent opera-

tions are carried out under water. Core internals are

removed and stored under water. The spent fuel is

then removed and placed in special underwater storage

racks within the reactor building. New fuel is then

419

installed into the core and the previous steps are re-

versed to place the reactor in operation again.

Under normal conditions and operations, no radio-

activity will escape. The entire refueling area is sur-

rounded by the reactor building which is automatically

isolated should an incident oceur which results in the

release of any radioactive material. Section 14.6.4 of

the FSAR, discusses the postulated radioactive re-

lease in the event of a refueling accident.

b.) Entire sealed fuel assemblies are removed during

refueling. The chemical composition of the fuel is

basically unchanged, that is an oxide, which now con-

tains fission products which are listed in d.) below.

ce.) No processing takes place at Vermont Yankee and no

toxie chemical compounds are present.

d.) Attached are two lists which present the significant

radioisotopes one day after shutdown, which is the

earliest that fuel can possibly be moved. One list is

of fission products and the second is an estimate of

other significant radio-nuclides present in a single

spent fuel assembly. The normal refueling procedure

removes only about 4% of a core during each shut-

down (about 92 assemblies).

aonourt Sie =

BR83

BR84

BR85

BR&7

1129

I31A

131B

1132

1133

1134

1135

1136

KR3M

KR5M

KR85

KR87

KR88

KR89

XEIM

XK3M

XE33

XE5M

XE35

XE37

X38

SES81

SKh3M

SES3

SES4

RBSs

RBs9

RB90

420

VERMONT YANKEE

FISSION PRODUCT INVENTORY

1 Spent Fuel Assembly

1 Day Decay

HALF-LIFE CURIES AFTER

(SEC.) 1 DAY DECAY

0,829K 04 0.158E 02

0.198E 04 0,.278E-08

0.180K 03 0.000E 00

0.563K 02 0,000K 00

0.505E 15 0.433E—02

0.695E 06 0.896E 05

0.695E 06 0.126 05

0.828E 04 0.1351 06

0.755E 05 O.115E 06

0.315E 04 0.895 E-02

0.241E 05 0.197 05

0.840E 02 0.000E 00

0.686 04 0.685E 02

0.158E 05 0.114KE 04

0.3346 09 0.167 04

0.468E 04 0.270E 00

0.100E 05 0.356K 03

0.192E 03 0.000E 00

0.108E 07 0.6441 03

0.198K 06 0.559K 04

0.459 06 0.244E 06

0.935E 03 O.59RE 04

0.326K 05 0.293KE 04

0,234E 03 0.000K 00

O.101K 04 0.689 E-20

O.108K 04 0,134E-19

0.6726 02 0.000E 00

O.150K 04 0.387 K-13

O.119K 038 0.000E 00

0.108E 04 0.401E 03

0.900E 03 0.290E-23

O.161E 03 0.000E 00

0.361 E-05

0.635B-15

0.000E 00

0.990E-09

0.204E-01

0,290 K-02

0,309 K-01

0.264E-01

0.204K-08

0.451K-02

0.000E 00

0.156K-04

0.260E-03

0.383 E-03

0.617 E-07

0813-04

0,000 00

0147-03

0.127K-02

0.559K-01

0.136102

0.671 E-03

0,000E 00

0.157 6-26

0.307 E-26

0.000E 00

0.886 K-20

0,000E 00

0.916E-04

0.663 E-30

0.000E 00

NUMBER ISOTOPE

RB91

RB92

SR89

SR9IO

SR91

SR92

SRI

SR94

Y90

Y9IM

Y92

Y93

Y94

Y95

ZR95

ZR97

NBOM

NB95

NB7M

NB97

MO99

M001

MO02

M005

TCOM

TOOL

TC2A

TC2B

TCOS

RUO3

RUO5

RU06

RUOT

RHSM

RH5M

RHO5

RELO6

RIH07

0.840E

0.800 E

0.441E

0.8841

0,350K

0.972E

0.420K

O.119K

0,.230K

0.300

0.529K

0.129K

0.359E

0.990E

0.630K

0.563K

0.613K

0.3231

0.302K

0.597

0.44418

0.244

0.900K

0.6931

0.1191

0.215K

0.8578

0.498

0.2691

0.597K

0.3445

0.161K

0.315K

0.287

0.341E

0.450K

0.1268

0.300E

O.131E

CURIES AFTER

1 DAY DECAY

0.231 K-25

0.000K 00

OAT9K 06

O.145K 05

0.4056 05

0.430E 03

0.000K 00

0.1458 05

0.2566 05

0.223K 06

0.743E 04

0.444E 05

0.125K-20

0.000K 00

0.2346 06

0.848E 05

O.457K 04

0.2441

0.8346

0,929K

0.184

ESSS

0.477 K-32

0.000K 00

O75 06

0.502 6-22

0.480K-32

0.781 K-32

0.000E 00

O112K 06

O.883K 03

0.128 05

0,000K 00

O112K 06

0.853E 03

0.227K 05

0.1296 05

O0.169B-15

0.528E-32

0.000E 00

0.409 E-01

0.332 B02

0.926602

0.982K-04

0,.000E 00

0,000 00

0.332 6-02

0.586 6-02

0.511B-01

0.16902

0.101E-01

0.28727

0,000K 00

0.5385K-01

0.193E-01

0.104602

0.558E-01

0.190KE-01

0,.212K-01

0.421K-01

0.558E-30

0.0001 00

0,000E 00

0.401K-01

0.114K-28

0,000E 00

0.178E-—38

0,000K 00

0.257K-01

0.201 K-08

0.292 E-02

0,000K 00

0.257E-01

0.194E-03

0.521 K-02

0,.295E-02

0,388 6-22

101

102

103

105

107

110

SN27

SN28

SN30

SB27

SB28

SB29

SB30

SB31

SB32

SB33

TE7M

TETA

TE7B

TE9M

TERIA

TE9B

TEIM

TEIA

TEIB

TH32

TE3M

TE33

TEH34

CS37

CS38

CS39

CS40

CS42

BATM

BA39

BA40

BA4I

BA42

LA40

LA41

LA42

LA43

CE41

CE43

422

HALF-LIFE

(SEC.)

O.741E 04

0.341 04

O.156E 03

O.319K 06

0.597TE 03

O.151E 05

0.597E 03

0.138E 04

0.126E 03

0.245E 03

0.907E 07

0.338E 05

O.341E 05

0.285F 07

O.372E 04

0.372E 04

0.103E 06

0.150E 04

0.280E 06

0.378KE 04

0.120E 03

0.263E 04

0.946E 09

0.192E 04

0.568E 03

0.660E 02

0.597E 02

0.156E 03

0.509K 04

O.110K 07

O.107E 04

0.660E 03

0.144E 06

0.136E 05

0.509E 04

O.114E 04

0.276E 07

O.11SE 06

CURIES AFTER

1 DAY DECAY

0.130E 01

0.340E-03

0.000E 00

0.443E 04

0.111 B-04

0.729E 03

0.15602

0.164E-13

0,000E 00

0.000F 00

0.133E 04

0.380E 04

0.797E 03

0.131KF 05

0.532E 03

0.130E 05

0.954E 04

0.469E-11

0.217E 04

0.131E 06

0.254E-01

0.128E-01

0.359E-04

0.144E 05

0.118E-07

0.000E 00

0.000F 00

0.135E 05

0.217E 01

0.227E 06

0.189E-18

0.000E 00

0.239E 06

0.316E 04

0.199E 01

0.352E-17

0.228E 06

O.137E 06

0.297 E-06

0.779 E-.0

0.000E 00

0.101 E-02

0.255E-11

0.166 E-03

0.357 E-09

0.376E-20

0.000E 00

0.304 E-03

0.870 E-03

0.182E-03

0.300 E-02

0.121E-03

0.299E-02

0.218E-02

0.107E-17

0.496 E-03

0.300E-01

0.580E-08

0.293E-08

0.821E-11

0.331 E-02

0.270E-14

0.000E 00

0.309 E—02

0.498 E-06

0.520E-01

0.432E-25

0.000E 00

0.547E-01

0.722 E-03

0.456E-06

0.804E-24

0.523E-01

0.314E-01

111 CE44

112 CE45

113 CE46

114 PR43

115 PR44

116 PR45

117 PR46

118 ND47

119 ND49

120 ND51

121 PM47

122 PM49

123 PM51

124 SM51

125 SM53

ISOTOPE

U-235

U-238

Pu-239

Pu-240

Pu-241

423

HALF-LIFE

(SEC.)

0.246E 08

0.180E 0°

0.840E 03

0.120E 07

0.105E 04

0.215E 05

O.144E 04

0.100E 07

O.719E O04

0.900E 03

0.794E 08

0.194E 06

0.100E 06

0.230E 10

0.169E 06

1 DAY DECAY

0.187E 06

0.000E 00

0.121E-25

0.222E 06

0.187E 06

0.937E 04

0.241 E-12

O.787E 05

0.966E O01

0.205 E-24

0.342E 05

0.302E 05

0.913E 04

0.110E 02

0.423E 04

TOTAL = 0.4375E 07

VERMONT YANKEE

Other Significant Radionuclides In

1 Spent Fuel Assembly

HALF LIFE

(Yrs)

7.1 X 10°

4.51 X 10°

2.44 < 10°

6.60 < 10°

1.43 X 10°

RATIO TO

TOTAL

0.427E-01

0.000E 00

0.277 E-32

0.507E-01

0.428 E-01

0.214E-02

0.551 K-19

0.179E-01

0.220 E-05

0.469E-31

0.782 E-02

0.691 E-02

0.208 K-02

0.252 E-05

0.966 E-03

MASS

(Kg)

1.60

185.0

0.84

0.45

0.15

The total activity of these radionuclides is about 2 < 10*

curies.

424

QUESTION 11.2

What impact would the chemical compounds described

above, in the quantities which will be present at the time

of refueling, have on the environment, including human

beings, if they were to escape the containment vessel?

ANSWER II.2

As no chemical compounds are employed, no environ-

mental effects are involved.

QUESTION 11.3

What impact would the radioisotopes present in the

materials removed from the reactor*¢ore vessel have on

the environment, including human beings, if they were to

escape the containment vessel? a.) Include short and long

term effects, specifying the length of time each isotope

would be dangerous to life. b.) Describe how these materials

might be disseminated through the environment — for ex-

ample, could they be disseminated through underground

aquifers at the plant site, or by the Connecticut River, or

by other means?

ANSWER IL3

Any release of these radioisotopes is considered unlikely

since they are contained in sealed fuel tubes and no pro-

cessing of the fuel is to be performed at the site. The only

postulated incident involves the dropping of a fuel as-

sembly while it is being moved from the reactor vessel to

the storage pool. This operation is performed under water

and within the confines of the reactor building. In the event

the fuel assembly were dropped and significantly damaged,

some of the fission product gases contained in the gap

between the fuel pellets and the cladding might escape and

rise to the surface of the storage pool. The analysis of

this incident is presented in Section 14.6 of the FSAR.

425

QUESTION 11.4

What facilities have been built into the plant to prevent

an accident which would disseminate the chemical and/or

radiological materials into the environment.

ANSWER I1.4

As stated in the answer to question II.2, there are no

chemical compounds employed in the nuclear coolant sys-

tem, therefore, none can be released in the event of an

accident.

Section 4+ of the station Final Safety Analysis Report

(FSAR) describes the reactor coolant system in detail.

The reactor vessel and connecting piping systems which

makeup the reactor coolant system provide a_ barrier

against leakage of radiological materials into the primary

containment. The reactor vessel is designed, fabricated, in-

spected and tested in accordance with the ASME Boiler

and Pressure Vessel Code, Section III. Reactor coolant

piping is designed and constructed in accordance with the

USA Standard Code for Pressure Piping, Power Piping,

USAS B 31.1.0, 1967. These codes require large safety

margins between the design pressure for vessels and piping

and the actual pressure where the material would fail.

Other requirements providing assurance of the reactor

coolant system barrier integrity are the quality assurance

and quality control programs employed during construc-

tion. In addition, an inservice inspection program will

affirm continuing integrity of the coolant barrier throughout

the life of the plans.

Section 7.2 of the FSAR describes the reactor protection

system. This system is designed to provide protection

against the onset of conditions that threaten the integrities

of the fuel cladding and the reactor coolant system barrier.

426

The system continuously senses the pertinent reactor para-

meters and will automatically perform its safety function

independent of operator action.

Reactor safety is further assured by operators who are

skilled, highly trained men who must demonstrate their

proficiency to the AEC Division of Reactor Licensing to

qualify for their positions. The operator is trained to

recognize conditions that could conceivably lead to a situa-

tion which would require action by the reactor protection

system and to take action to safely shutdown the reactor.

In short, many considerations are taken into account in

the design, fabrication, construction, testing and operation

of a reactor plant to provide a high degree of certainty that

an accident will not occur.

QUESTION IL5

Assuming that the chemical and/or radiological ma-

terials, or some part thereof, did escape the containment

vessel, what plans exist for preventing further dissemina-

tion of them into the environment?

ANSWER IL5

This response is directed to the radiological aspects of

the question since no chemical compounds are present in

the reactor coolant system.

There are numerous station facilities built for the pur-

pose of preventing dissemination of radiological materials

into the environs in the event of an accident. Detailed de-

scriptions of these facilities and how they function during

accident conditions are a substantial part of the written

work presented in the four volumes of the plant FSAR.

Section 5 of the FSAR describes the primary and sec-

ondary containments and the functions of each to mitigate

427

the effects of an accident. Section 14.6 of the FSAR pre-

sents analyses of various design basis accidents including

responses of the primary and secondary containments to

the design basis accidents where applicable.

Section 6 of the FSAR describes the systems (core

standby cooling systems) which will automatically provide

sufficient cooling to the reactor core in the event of a loss of

coolant accident. Analyses for the various loss of coolant

accidents are also presented in Section 6.

Section 7 of the FSAR describes the instrumentation and

controls used for reactor protection circuitry, primary and

secondary containment isolation circuitry, reactor core iso-

lation cooling systems circuitry as well as other reactor

control systems.

QUESTION I1.6

Under what circumstances is it credible that the chemical

and/or radiological materials described in the above sched-

ule of questions, or any part thereof, could escape from the

containment vessel? Supply any relevant studies upon

which this judgment is based.

ANSWER I1.6

As mentioned previously in answers II.2, I1.4, and IT.5,

no chemical compounds are employed, therefore, the por-

tion of the question pertaining to chemicals is not applic-

able.

Section 14.6 of the FSAR presents detailed analyses of

the credible design basis accidents and the analytical re-

sults in terms of off site doses to the public. A credible

design basis accident is a hypothesized accident the char-

acteristics and consequences of which are utilized in the

design of those systems and components pertinent to the

preservation of radioactive material barriers and the re-

428

striction of radioactive material release from the barriers.

The hypothesized accident is a “worst case” situation

which, even though it is extremely unlikely ever to happen,

is assumed to occur coincident with other highly unlikely

events such as loss of off site electrical power and an earth-

quake. The doses are calculated based upon radiological

materials escaping into the environs. The quantity of

escaping radioactive materials are analytically determined

based on a series of very conservative assumptions (con-

servative in that actual dose rates would likely be much

less than those shown in the analysis).

Section 8 of the FSAR describes emergency electrical

power circuits, components and sources (AC and DC).

Ill. FUEL HANDLING

QUESTION IIl.1

Describe the procedures by which fuel materials received

at the plant gate will be processed and stored, specifying

the accounting and security procedures which will be used

to prevent accidental loss or theft of these materials during

receiving, storage prior to loading, or while the fuel is be-

ing loaded into the reactor.

ANSWER III.1

The following procedure provides the answer to this

question.

1.0 SCOPE

This procedure establishes requirements for the control

of special nuclear material in the facility of Vermont

Yankee Nuclear Power Corporation where such material is

used directly or indirectly in the production of electrical

energy. The materials involved in this facility are uran-

ium and plutonium of various isotopic content. The nuclear

429

material control system involves the establishing of re-

ceipts from the shipper, six internal material balance areas,

and records of shipments to a reprocessor or other licensed

receiver as the means of maintaining control and account-

ability of all leased, or owned special nuclear material, or

special nuclear material owned by others but residing in

the facility of Vermont Yankee.

2.0 DEFINITIONS

2.1 Special Nuclear Material (SNM)

A. 10CFR 70

Special Nuclear Material as defined by 10 CFR

70 is as follows: “(1) Plutonium, Uranium

-233, uranium enriched in the isotope 233, or

in the isotope 235, and any other material

which the Commission, pursuant to the pro-

visions of Section 51 of the act, determines to

be Special Nuclear Material, but does not in-

elude source material, or; (2) any material

artificially enriched by any of the foregoing

but does not include source material.”

B. Nuclear Material Control and Safeguards

System For the purposes of these proceedings

the following shall be considered to be units

of special nuclear material:

1. Fuel Assemblies or separate fuel pins

2. Source Range Monitors (SRM’s)

3. Intermediate Range Monitors (IRM’s)

4. Local Power Range Monitors (LPRM’s)

5. Traversing In-Core Probe Monitors

(TIP’s)

6. Subcritical Training Assembly

2.2

2.3

3.1

430

Material Balance Areas

For the purpose of maintaining control over the

special nuclear material at the reactor site, the

plant shall be divided into the following special

nuclear material balance areas:

A. New Fuel Storage Vault

B. Spent Fuel Storage Pool

C. Reactor

D. Reactor Operating Floor

KE. Warehouse

F. Training Laboratory

Authority and Responsibility

All special nuclear material on site shall be under

control of the Reactor Engineer by direction of

the Plant Superintendent. All reporting to the

Central Accountability Office shall be done through

the Reactor Engineer.

3.0 RECEIPT OF SPECIAL NUCLEAR MATERIAL

Kach receipt of special nuclear material shall be

accompanied by a bill of lading or packing list

which contains a listing of the shipping containers

involved in the shipment as well as a loading sched-

ule for each shipping container. The loading

schedule should show:

(a) The serial number of the shipping container

(b) The serial number of the individual units of

special nuclear material contained in the ship-

ping container.

431

3.2 A piece count of shipping containers shall be per-

3.3

formed upon receipt and the serial numbers of the

shipping containers received shall be compared

with those shown on the shipping documents.

Upon removal of the units of special nuclear ma-

terial from the shipping containers, the serial num-

bers of the individual units of special nuclear ma-

terial received shall be compared with those shown

in the shipping documents.

After the unloading of each shipment the Central

Accountability Office shall be notified of the date of

receipt of the special nuclear material and the

serial numbers of the units of special nuclear ma-

terial using a pre-printed form, VY0002.

40 STORAGE OF SPECIAL NUCLEAR MATERIAL

4.1

4.2

4.3

The storage of all receipts of special nuclear ma-

terial shall be in a Material Balance Area. Opera-

tions pertaining thereto shall be restricted to

selected employees named by the Reactor Engineer

and approved by the Plant Superintendent. There

is no personal access to the new fuel vault.

A record shall be maintained of the contents of all

storage locations showing the serial numbers of all

units of special nuclear material received or re-

moved, the dates of such receipts or removals, and

the location of each unit of special nuclear material

in the storage location.

Immediately upon removal of units of special nu-

clear material from their storage locations the

Central Accountability Office shall be notified of

the date of such removal, the serial numbers of the

units of special nuclear material removed, and the

disposition of the removed units,— using pre-

printed forms, VY0003, and VY0004, VY0005 or

V Y0006.

5.0 REACTOR OPERATION

5.1

5.2

5.3

5.4

A unit of special nuclear material is introduced in-

to or removed from the Reactor only during re-

fueling.

The Central Accountability Office shall be promptly

notified of the location of each unit of special nu-

clear material in the Reactor upon its introduction

thereto, using VY0003 and VY0007 or V Y0008 and

of any relocation of units of special nuclear mater-

ial during refueling operations using VY0007 or

VY0008.

A record shall be maintained in the Reactor Con-

trol Room showing the serial numbers of the units

of special nuciear material charged to or removed

from the Reactor, the dates of such charging or

removal, and the location of each unit of special

nuclear material in the Reactor. At the conclusion

of refueling the Central Accountability Office shall

be furnished with a copy of the Core Fuel As-

sembly lay-out plan, VY0007, and in-core instru-

mentation lay-out form VYOOO8.

Immediately upon removal of irradiated units of

special nuclear material from the Reactor, the

Central Accountability Office shall be notified of

the date of such removal, the serial numbers of the

units of special nuclear material removed, and the

disposition of the removed units using preprinted

forms V Y00038 and V Y0009.

433

5.5 A member of the Plant Reactor Engineers staff at

the Vermont Yankee site shall be responsible for

the calculation of Core depletion and plutonium

production. He shall also determine the place-

ment of assemblies in the reactor at the time of

refueling.

At the end of Core life the Central Accountability

Office shall be notified of the depletion of special

nuclear material by units in both total uranium

and Uranium 235, and the plutonium production

the Core has experienced during its residence in

the Reactor, identifying each unit by serial num-

ber. During Core life at the end of each reporting

period, using form AO-1, the Central Accounta-

bility Office shall be notified of the uranium deple-

tion and plutonium production experienced in the

Core during the period to be reported.

6.0 STORAGE OF IRRADIATED SPECIAL

NUCLEAR MATERIAL

6.1

6.2

6.3

The storage of all irradiated units of special nu-

clear material shall be effected in a Spent Fuel

Pool, which will be located in a controlled Access

Area. Access to the area shall be restricted to

selected employees.

The Central Accountability Office shall be notified

of the location of each irradiated unit of special

nuclear material in the Spent Fuel Pool upon its

receipt, and of any subsequent change in location

using forms V Y0003.

A record shall be maintained in the Spent Fuel

Pool area showing the serial numbers of the units

of special nuclear material received at and removed

from the area, the dates of receipt and removal

6.4

434

f rom the Spent Fuel Pool, and the location of each

unit of special nuclear material. The location of

the units of special nuclear material in the Spent

Fuel Pool shall be recorded on VY0009 and a copy

of this form shall be sent to the Ce

7k » Central A .

ability Office, ecount

Immediately upon removal of irradiated units of

special nuclear material from the Spent Fuel Pool

the Central Accountability Office shall be notified

of the date of such removal, the serial numbers ot

the units of special nuclear material removed, and

the disposition of the removed units i

| — using a pre-

printed form VY0010. tien

7.0 SHIPMENT OF SPECIAL NUCLEAR MATERIAL

7.1

7.2

Hach shipment of special nuclear material shall be

accompanied by a bill of lading or packing list

which shows the serial number of the shipping cask

involved and a loading schedule which shows the

serial numbers of the individual units of special

nuclear material contained in the eask.

After the loading of each shipment, the Central

Accountability Office shall be notified of the date of

shipment of special nuclear material, the serial

numbers of the units of special nuclear material

shipped, the serial number of the shipping cask

ea Bo consignee of the shipment using form

8.0 INVENTORY OF SPECIAL NUCLEAR

MATERTAL

8.1

At least semi-annually a complete inventory of all

new, irradiated, or Reactor contained special nu-

clear material shall be taken.

8.2

8.3

8.4

8.5

8.6

435

A physical inventory of new and irradiated units

of special nuclear material shall be taken and their

individual serial numbers recorded. The identity

of units of special nuclear material contained in the

Reactor Core shall be taken from Control Room

records.

The special nuclear material content (isotope and

element) of new units of special nuclear material

shall be assumed to be that represented to be con-

tained in each unit by the fabricator and which

may be subject to inspection verfication.

The Central Accountability Office shall be notified

by the site staff of the depletion of special nuclear

material stated in both total uranium and Uranium

235 and plutonium production. The values so re-

ported shall constitute the special nuclear material

inventory of the Reactor Core.

The special nuclear material content (isotope and

element) of irradiated units of special nuclear mate-

rial shall be assumed to be that as computed and

submitted to the Central Accountability Office as

described in Section 5.6.

The limit of error in determining (or estimating)

the amount of special nuclear material contained

in the units of special nuclear material comprising

the inventory shall be established and the overall

measurement uncertainty for the inventory shall

be determined by propagation of error techniques.

9.0 RECORDS AND REPORTS

91 The Central Accountability Office shall maintain

records of receipts, transfers and shipments of spe-

9.2

9.3

436

cial nuclear materials as defined in Section 2.1B

items 1 and 6 in each material balance area. These

records shall show serial numbers for each unit of

special nuclear material and the weight of con-

tained special nuclear material (element and iso-

tope), the composite limit of error associated with

the determination of the latter, and a reference for

the determination of the limit of error involved

(such as the fuel fabricator’s shipping document

and the Reactor Engineer’s calculated burn-up).

These records shall be organized such that a run-

ning inventory of special nuclear material can be

determined readily from the records.

The Central Accountability Office shall maintain

records of receipts, transfers, and shipments of

special nuclear materials held off-site by other

licensees for Vermont Yankee is financially re-

sponsible for these materials. The Central Ac-

countability Office shall also maintain records of

materials owned by others but residing in the facil-

ity of Vermont Yankee. These records shall show

the same information as set forth in (1), above.

The fabricator of units of special nuclear material

shall provide a special nuclear material transfer

document to the Central Accountability Office for

each shipment of special nuclear material to Ver-

mont Yankee. This transfer document shall con-

tain, or shall have attached thereto, and made a

part thereof, supplementary data sheets showing:

(a) Unit of special nuclear material identification

numbers.

(b) Weight of special nuclear material (element

and isotope) contained in each fuel assembly.

9.4

9.5

437

the determination of the special nuclear ma-

terial content of the material contained.

Additionally, it is desireable that the fabricator

provide for Vermont Yankee, and Vermont Yankee

will try to obtain the following information from

the fabricator for units of special nuclear material

furnished as fuel assemblies:

1) Fuel rod identification numbers for each fuel

assembly.

2) Weight and lot number of fuel material in

~se each fuel rod.

3) Weight of special nuclear material (element

and isotope) contained in each fuel rod.

4) A correlation between the UQ: pellet lots and

the fuel rod numbers wherein the pellets are

contained, and the results of special nuclear

material analysis of each UQOz pellet lot (ele-

ment and isotope).

The documentation described in Section 3.3 con-

cerning receipts of units of special nuclear material

shall be checked against the special nuclear mate-

rial transfer document provided by the shipping

licensee. Discrepancies shall be investigated and

resolved promptly.

The transfer document and the stated special nu-

clear material content of the units of special nu-

clear material shall be certified by a responsible

member of the shipping licensee’s management

staff. The shipping licensee’s values of the special

nuclear material content units of special nuclear

material shall be entered in the Accountability

Ledger for Vermont Yankee on an individual basis.

458

9.6 Two Control Accounts shall be maintained showing

9.7

9.8

9.9

total amount of special nuclear materials held on

lease or owned. One Control Account shall show

total uranium, Uranium 235, and total dollar valu-

ation of the material leased or owned. The second

Control Account shall show total plutonium total

Plutonium 239 plus Plutonium 241, and dollar val-

uation of the material leased or owned.

The Central Accountability Office shall prepare a

special nuclear material transfer document for

each shipment of special nuclear material. This

transfer document shall have attached thereto, and

made a part thereof, VY0010 supplementary data

showing:

(a) Unit of special nuclear material identification

numbers.

(b) Estimated weight of special nuclear material

(element and isotope) contained in each unit

of special nuclear material as obtained from

burn-up ealeulations described in Section 5.6.

involved in the burn-up calculations.

The values shown on the transfer document for the

special nuclear material content of the units of

special nuclear material shipped shall be entered

in the Accountability Ledger for the Power Re-

actor Facility on an individual shipment basis.

Upon receipt of the reprocessor’s report of meas-

urements of the special nuclear material content of

the irradiated units of special nuclear material

during reprocessing, and approval thereof, the

Accountability Ledger shall be adjusted to reflect

the results of the reprocessor’s measurements.

439

9.10 The Central Accountability Office shall prepare a

material balance for special nuclear materials for

each six month period showing beginning inven-

tory, receipts, shipments, burnup, ending inventory

and the uncertainty connected with each.

QUESTION III1.2

How long does the process of loading and reloading take?

ANSWER IIL.2

Loading and reloading is defined as the time required to

remove designated used fuel from the reactor, relocate par-

tially used fuel to new core positions and install new fuel

into the reactor. The stated times do not include time for

preparation and completion of loading such as reactor

vessel head removal and replacement.

The process for loading and reloading the fuel initially

will require twelve days. On subsequent reloadings after

the plant has been operational, the twelve days will be re-

duced to approximately seven days. This reduction in time

required on the subsequent reloadings is due primarily to

the fact that some work can be accomplished prior to the

refueling. Channeling, for instance, which consists of in-

stalling a fuel assembly into a shroud-like enclosure, is per-

formed on a partial number of fuel assemblies prior to the

actual loading operation.

QUESTION II1.3

Ilow many workers are involved in this process?

ANSWER II1L.3

The total number of workers involved in this process are

twenty-one. This is broken down into seven per shift. The

operation will involve around the clock work consisting of

three shifts.

440

QUESTION IIl.4

What precautions are taken in the selection of personnel

to handle this task to assure that none will steal or lose fuel

materials during receiving, storage, or loading of fuel

materials?

ANSWER III.4

All applicants for plant staffing are screened and inter-

viewed in a manner that would disclose if any person is

mentally disturbed. All applicants meet and are inter-

viewed by a minimum of three plant supervisory staff mem-

bers, who explore applicants background history and pres-

ent and future goals.

QUESTION 11.5

What precautions will be taken to assure that no nuclear

material will be removed from the plant grounds by workers

at the plant?

ANSWER IIL.5

A Station Security Program has been established to pro-

vide physical security, protection against impairment of

necessary plant functions, and theft of Company equipment.

A Security Force is established and is responsible to achieve

these goals. The detailed Industrial Security Plan is con-

sidered to be proprietary since divulging its contents to the

public would seriously impair its effectiveness.

The physical size and weight of the fuel is a major factor

in preventing its removal from the plant site. Each as-

sembly is approximately 14.5 feet long, 5.4” x 5.4” square

and weighs approximately 580° pounds. Once fuel has been

irradiated in the reactor, it has to be handled under water

1 This is a typographical error in the original and should read “680”

instead of “580”. See FSAR (Ez. 1) §1.7, Table 1-7-1.

441

or in a heavily shielded cask. Such a cask would weigh

many tons, requiring the use of a large crane to handle it.

QUESTION II1.6

What is the percentage of nuclear fuel material expected

to be lost, stolen, or otherwise unaccounted for at the plant

site over the expected life of the plant, based upon previous

experience of the Atomic Energy Commission?

ANSWER III.6

The percentage of nuclear fuel material expected to be

lost, stolen, or otherwise unaccounted for at the plant site

over the expected life of the plant is zero. Since we are

unable to speak for the Atomic Energy Commission, this

statement is based upon eleven years of operation at the

Yankee Rowe Plant and four years of operation at the

Connecticut Yankee Plant.

QUESTION 111.7

What would be the maximum impact upon the environ-

ment, including persons, should this amount of material

find its way into the environment outside the plant grounds?

ANSWER III.7

Since zero loss of nuclear material is expected, there

would be no impact upon the environment.

QUESTION IIL8

How much of the material expected to become lost, stolen,

or unaccounted for at the plant site during the life of the

plant, is expected to find its way into the environment, based

on the past experience of the Atomic Energy Commission?

ANSWER IIL8

Since none of the material is expected to become lost,

stolen, or unaccounted for at the plant site during the life

442

of the plant, none is expected to find its way into the en-

vironment.

QUESTION I1.9

Supply any relevant studies upon which the judgments

called for above are based.

ANSWER IIL9

The above judgments are based on a total operating ex-

perience of fifteen years. This operating experience consists

of eleven years at the Yankee Nuclear Power Plant in Rowe,

Mass. and four years at the Connecticut Yankee Haddam

Neck Plant. During five of the years at the Yankee Rowe

Plant, per agreement with the U. S. Government, the Plant

was under voluntary inspection performed by the Inter-

national Atomic Energy Agency. The purpose of these

inspections was to verify that no nuclear material had been

lost, stolen, or diverted from the Plant. It should be noted

that the Yankee Rowe Plant was the only nuclear plant

whose management volunteered to submit to such periodic

inspections. The results of these inspections again verified

that no nuclear material was lost, stolen, or unaccounted

for at the Plant site.

Vv. TRANSPORTATION — FUEL FOR THE PLANT

Comment: New fuel shipped to the plant is in the form

of fuel assemblies. The uranium oxide and mixed uranium-

plutonium oxide is in the form of high density, ceramic pel-

lets contained in sealed Zircaloy tubes which together with

other hardware are assembled into fuel assemblies.

1. Q. Describe the package in which fuel will be delivered

to the Vermont Yankee plant. Include information as

to its weight, size, and markings.

443

A. The packages used for delivering fuel must be spe-

cially-designed for the type of fuel assembly and spe-

cially approved by the Atomic Energy Commission

(AEC) and the Department of Transportation (DOT)

prior to their use. A description of the package being

used for Vermont Yankee’s first fuel together with

weight and size information is contained in Exhibit

V-A. The package is marked in compliance with DOT

regulations for shipping purposes as follows:

Consignee’s name and address

Shipper’s name and address

e. DOT Special permit number

d. Type of package

e. Transport Group number

f. Fissile Class number

g. Radioactive Material label

h. CG mark (Center of Gravity)

The package(s) to be used for subsequent fuel assem-

blies is unknown; however, it will comply with AEC

and DOT requirements. The radioactivity level at the

surface of the package is approximately 1 mr/hr.

Q. What amount of radioactive fuel material will be

contained in each package? Express this in terms of

mass — e.g. in kilograms — and in terms of its chem-

ical composition and radioactivity (in appropriate

units).

A. The fuel assemblies in each package contain ap-

proximately 440 kilograms of uranium dioxide (387

kilograms of uranium) and will have no more than 0.5

euries of radioactive material.

444

3. Q. What percentage of the fuel to be used at the Ver-

mont Yankee plant over its useful life will be pluton-

ium: Give the information asked for in V.1 and 2

above with respect to this fuel material.

A. It is planned to include in the first fuel used, four

assemblies that are partially fueled with plutonium for

the purpose of obtaining technical data. These assem-

blies are nearly identical in design to all-uranium

assemblies, the major difference being that some of the

sealed rods contain uranium-plutonium mixed oxide

pellets rather than all uranium oxide pellets. The plu-

tonium in these assemblies will comprise about 0.009%

of the combined uranium and plutonium in the initial

fuel loading. At some time in the future it is probable,

after having obtained the required regulatory ap-

provals, that plutonium will be used to replace some of

the uranium in new fuel assemblies. If it is assumed

that all plutonium produced in Vermont Yankee is used

in new fuel as early as possible and is continued to be

used until fast breeder reactors require plutonium for

fuel, the schedule for new fuel insertion and its uran-

ium-plutonium composition might be as follows:

Percentage

Year of Plan lomposition o

@nane tion When . New Fuel .

New Fuel is

Inserted Uranium Plutonium

0 99.991 0.009

1- 4 100 0

5-19 99.3-99.0 0.7-1.0

20-30 100 0

Therefore, it is expected that no appreciable quantity

of plutonium will be used in new fuel until after 45

years of plant operation and that it will then be used

445

for a period of about 15 years. The package to be used

for shipping the four part-plutonium assemblies will

be the same as that used for all-uranium assemblies as

described for Question V.1. The package to be used for

future part-plutonium assemblies is unknown. All

packages and shipments will comply with AEC and

DOT requirements. The mass of material will be about

the same as in V.2, above. The chemical composition

of material in the sealed Zirealoy rods will be a mixture

of uranium dioxide and plutonium dioxide. If future

part-plutonium assemblies are similar in design to the

first four the amount of radioactivity per container is

estimated to be about 18,000 curies. Measurements on

rods making up the first part-plutonium assemblies in-

dicate a gamma activity of only 2 mr at a distance of 2

feet from the rods and 12 mr at a distance of one inch.

@. How many packages of uranium fuel will be deliv-

ered on each truck/train (whichever will be used for

delivery — both if both apply)?

A. The first fuel is being shipped by truck. Each

truckload consists of up to 16 containers, each contain-

ing two fuel assemblies. It is not known whether future

new fuel shipments will be by truck or by rail. Facili-

ties for receiving both types of shipments exists at the

VY plant. It is not known how many containers will,

in the future, be loaded onto each vehicle. However,

all shipments made will be in compliance with AEC and

DOT regulations which specify limits for the number of

containers on a vehicle.

. Q. Where will this fuel be shipped from?

A. All of the first fuel assemblies (710,840 kg uran-

ium) and 2.4 reloads of fuel assemblies (55,440 kg

446

uranium) are shipped from Wilmington, N. C. Since

the manufacturers of subsequent fuel reloads have not

been selected it is not known from what locations such

reloads will be shipped.

. Q. By what route will the fuel travel in order to reach

the Vermont Yankee plant? Give route numbers of

roads which will be used, specifying any cities through

which trucks will pass. If railroad is to be used, spe-

cify cities through which the trains will pass and indi-

cate clearly where these train routes are.

A. The route of first fuel shipments begins by taking

US-421 from Wilmington, N. C. to its junction with

Interstate-95; thence north on Interstate-95 to its june-

tion with US-301 in Richmond, Va. Take US-301

north, bypassing Washington, D. C. to east and cross

Chesapeake Bay via the Annapolis-Queenstown Bridge

to Wilmington, Delaware. Through Wilmington and

across the Delaware River over the Delaware Memorial

Bridge on Interstate-295. Travel 295 to its junction

with US-130. North on US-130 through Camden, N. J.

but bypassing Trenton, N. J. to the junction of US-130

and US-1 and 9. North on US-1 and 9 over the George

Washington Bridge in New York City to its junction

with Interstate-95. East on 95 through Bridgeport,

Conn. to New Haven where connection is made with

Interstate-91. Through New Haven north on Inter-

state-91 through the cities of Hartford, Conn., Spring-

field, Mass. and Brattleboro, Vt. At Brattleboro con-

nect with State Route-142 south to the plant site at

Vernon, Vt.

. Q. What, if any, calculations have been made to de-

termine the chances of an accident involving fuel-

447

carrying vehicles over the life of the plant? If such

calculations have been made, what are the results?

A. The National Safety Council publishes as public

information, yearly statisties of all highway accidents,

including as part of the report, truck accidents involv-

ing hazardous materials. These reports indicate there

are so few nuclear fuel truck accidents each year that

any statistic obtained by this method would not be of

worthwhile significance if we attempted to use it as a

life-of-the-plant vehicular accident probability factor.

Note, however that in the event that an accident did

occur involving nuclear fuel, simulated accident tests

which the fuel containers have successfully withstood

demonstrate that the fuel would remain intact, more-

over even if through some catastrophic event, the fuel

material (UO, pellets) were dispersed, no significant

environmental effect would occur.

. Q. What, if any, tests have been conducted using the

type of containers in which fuel elements are to be

shipped, to determine the conditions under which they

might burst open or otherwise release radioactive ma-

terial?

(a) Have any of these tests been conducted under

actual conditions — involving dropping the container

from a moving truck or train, for example — as Op-

posed to in a laboratory with standard bursting and

puncture tests?

(b) If such tests have been conducted, describe the

tests and the results.

A. The fuel package tests conducted are the water

immersion, fire and 30-foot drop tests. Specifications

and standards are established by the AEC as pre-

10.

448

seribed in LOCFR, Part 71: 31, 32, 34, 35, 36 and Ap-

pendices A & B and for the DOT as prescribed in

49CFR, Part 173:393 & 398. Results for first fuel

packages are documented in Exhibit V-A.

(a) No tests for first fuel packages have been

conducted under actual conditions involving dropping

the containers from a moving truck or train.

(b) See (a).

Y. Describe the methods used to secure the containers

in which fuel elements will be shipped to the Vermont

Yankee plant to the trucks or railroad cars on which

they will be shipped.

A. The method used to secure first fuel containers to

the truck is to pass 3¢” steel link chain up the sides of

the containers, over the top of the container block and

down the other side. Both ends of the chain are

hooked loosely to the trailer stake pockets. A chain

binder is then placed at one side of the chain to pick up

the slack in the chain. The slack is picked up one link

at a time until the chain is so tight that one man using

a length of steel pipe can scarcely close the binder.

Three chains are used on each container or container

block, spaced equi-disiantly from each other along the

length of the containers.

(. Approximately how long will be required to re-

lease the containers from these fasteners? How is this

accomplished ?

A. Releasing the containers is accomplished by open-

ing the chain binders. Normally, this requires very

little time — 15 minutes at most. Subsequent unload-

ing of the containers, however, is not so easily accomp-

11.

12.

13.

449

lished because of their weight (2800 pounds each) and

materials handling equipment is required.

(. How many stops will be made hciween the time the

truck(s) or train(s) leave the original shipping point

and the time they reach the Vermont Yankee plant?

A. All first fuel shipments are by truck. Normally

four to six stops are required by the truck from the

time it leaves Wilmington, N. C. until arrival at the

plant site at Vernon, Vt.

@. What is the value, in dollars, on the present Amer-

ican market of the quantity of fuel to be shipped on

each truck/train?

A. Each full truckload of 16 containers of the first fuel

will carry fuel assemblies valued at $1,800,000. Of this,

the value of the contained uranium is $1,000,000, the

balance of $800,000 being for fabrication of the fuel

assemblies. For the fuel assemblies as such, to have

the full value of $1,800,000 to another party on the

American market the other party would need to have a

reactor nearly identical in design to VY in which to use

them. For the contained uranium to have value to

others, in itself, it wonld be necessary to reprocess the

assemblies to recover the uranium in a usable form.

In addition to destroying the fabrication value, repro-

cessing could only be done at considerable expense

which would effectively reduce the value of the uranium

to others. If the uranium recovery were done in a

reprocessing plant the uranium value would be de-

ereased from $1,000,000 to about $800,000.

Q. What is its value on the international market?

A. On the present international market the situation

would be essentially the same as for the American mar-

14.

15.

450

ket (Question 12) except that the contained uranium

might be slightly less in value.

@. What studies have been made to determine the

likelihood of hijacking or theft of fuel or fuel contain-

ers during transit to the Vermont Yankee plant? Sup-

ply studies and describe their conclusions.

A. Although there have been no specific studies made

to determine the likelihood of hijacking or theft of fuel

during transit to Vermont Yankee, the manufacturer of

the first fuel assemblies has an established fuel moni-

toring system requiring each vehicle transporting fuel

to report its position and other important information

by telephone every six hours during the trip to a re-

sponsible monitoring unit. This frequent telephonic

surveillance of each shipment would quickly alert the

fuel manufacturer to any attempted hijack or theft of

nuclear fuel. Fuel shipments are made by direct sleep-

er truck-trailer service with two drivers and no trans-

fer of equipment or drivers straight through to desti-

nation. The size and weight of the fuel containers

would very likely preclude any theft or hijacking ex-

cept by use of heavy duty special purpose handling

equipment.

Q. What is the present best estimate of the chances

of a hijacking or theft of fuel during shipment to the

Vermont Yankee plant over the estimated life of the

plant, assuming presently planned precautions are

taken?

A. The present best estimate is that the chance of a

hijacking or theft occurring during shipment of Ver-

mont Yankee fuel is remote because of the precautions

taken during shipment, and lack of any illegal use for

16.

17.

18.

451

the material. To the best of our knowledge, there has

never been a theft or hijacking, or an attempt at theft

or hijacking, during the shipment of nuclear fuel as-

semblies.

@. What would be the likely environmental impact,

including impact on persons, of the release of radio-

activity from a shipping container which was stolen or

lost in transit, and later opened?

A. There is absolutely no environmental impact or

effect on living creatures upon contact with or proxim-

itv to intact nuclear fuel assemblies because of their

insignificantly low radiation level.

(). How many containers of plutonium or other fuel

would have to be stolen in order for someone who stole

them to assemble enough material to create a critical

mass?

A. If the fuel assemblies were removed from the con-

tainers and arranged closely in water criticality could

be achieved with assemblies from two containers (that

is 4 assemblies).

(. How many such containers will be shipped to the

Vermont Yankee power plant over its expected useful

life?

A. Assuming the plant will have a useful life of 30°

years it is estimated that 1700 such containers would

be shipped during that period.

1 This is a typographical error in the original and should read “40” in-

stead of “30”.

452

EXHIBIT V-A

Description of Model RA-1] Package Used For

Vermont Yankee’s First Fuel Assemblies

1.0 Package Description — Packaging

(a) General: A right rectanguwar box consisting of a

wooden outer container and a metal inner container

separated by cushioning material.

(b) Gross Weight: 2800 pounds.

outer shell and perforated inner basket separated

by struetural angle iron. Outer shell formed of

16-gange carbon steel plate with an integral welded

end of same material. Four 1%” thick one-inch

angle stiffeners are welded on 4” centers to the

outer surface of the end plate. Approximate di-

mensions 9” high, 18” wide and 174” long.

Inner basket of 16-gauge steel plate with 34” per-

forations on 134” centers welded to upper edge of

outer shell to form two U-shaped channels 67%”

square which may be lined with cemented low-

density cushioning with perforations matching size

and location of those in basket.

Inner basket supported within inner shell by four

1%” thick 3” x 3” angle iron spacers positioned

longitudinally along entire length of body. Cover

and end cap constructed similar to the box to pro-

vide 2” annulus around fuel when closed except at

ends. A manually operable pressure relief valve

passes 2 cfm air automatically at .5 psi pressure

difference. Gasket of 14” thick hollow neoprene

provides full seal with cover in place. Closure

453

effected by latches (“Camloe”, 3711 series or equiv-

alent) which are inaccessible during transport.

(ad) Outer Container: Box 33” high, 32” wide and up to

207” long fabricated of 14” plywood, 2 x 4 cleated,

on 2 x 6 tongue and groove platform with bolted

4 x 4 skids.

(e) Internal Cushioning: Three 3” layers of 34” ceil

phenolic resin impregnated kraft fiber honeyeomb

line the box at ends; one layer at top, bottom and

sides. Remaining space at ends filled with ex-

panded polyethylene cushioning. Three-inch-thick,

12” wide pads of same material centered over

transverse skids at sides, top and bottom. No

attached lifting or tiedown devices.

2.0 Package Evaluation

A sample Model RA-1 package was subjected to the

tests or assessments set forth in Subpart C of 10 CFR

71. Results of these tests and assessments were sub-

mitted for the AEC’s review as part of the fuel manu-

facturer’s application for an amendment to its special

nuclear materials license for authorization to use con-

tainer Model RA-1 for shipments of fuel assemblies.

This evaluation data is given below.

2.1 General

There are no components of the packaging or its

contents which are subject to chemical reaction in

normal transportation environment. The package

cannot be opened inadvertently, uses no coolant,

and has no lifting devices or tiedown attachments.

454

2.2 Normal Transport Conditions

(a)

(b)

(c)

(d)

Thermal —None of the components of the

fuel assemblies or the inner metal container

on which containment integrity and nuclear

safety depend are significantly affected by tem-

peratures within the range of —40°F to 130°F.

Pressure — The breather valve opens auto-

matically when the inner container is subjected

to =.5 psi pressure differential. Therefore

there is no effect on the packaging from an en-

vironment of .5 atmosphere.

Vibration— A 6” thick layer of cushioning

material surrounds the inner metal container

at the sides, top and bottom with an additional

3” thickness at the ends. The Ethafoam cush-

ioning is slightly compressed in final closure

and banding of the outer container and is

therefore not free to shift during transport.

Since the latches on the metal container are

pinned they cannot loosen during normal trans-

port vibration or shock even were such vibra-

tion able to penetrate the cushioning material.

Water Spray and Drop Tests — The complete

package is designed to protect the fuel as-

semblies within the inner metal container

from loss of containment integrity or change

in nuclear safety reliability by virtue of the

thick enshioning material surrounding it. The

shock absorbing qualities of this material and

of the heavy wooden cleats at the corners,

edges and at all joints in the plywood, supple-

mented by the inherent resiliency of nails and

(e)

(f)

(g)

455

steel bands used in final closure of the outer

package constitute a more than adequate buf-

fer against the subject tests. This conclusion

is further supported by analysis of high speed

motion pictures recorded during the 30-ft

drop test, described in Section 3.0, in which

the flexibility of the package under impact is

clearly evident.

Penetration — Tests were conducted in which

the flat circular end of a vertical steel cylinder

1144” in diameter weighing 13 lbs. was dropped

four feet onto the center of the 4%” plywood

outer container. No damage resulted after

four drop tests.

Compression — Tests were conducted in which

six loaded packages were stacked. There was

no visible nor measurable damage to the con-

tainer on the bottom of the stack.

Reactivity—In the normally undamaged

condition, fuel assemblies are dry. Con-

sideration was given to the effect of the hydro-

genous outer packaging materials between ad-

jacent metal containers. However, the analysis

of water immersion of an infinite array of

packages without water inside the inner con-

tainer, described in detail in Section 3.1 below,

demonstrates that condition would produce a

more reactive array than the as-shipped con-

dition. Results of conservative calculations of

the as-shipped condition indicate a value for

ko of 0.8. This seemingly high value is par-

tially explained by the degree of moderation of

the packaging material and partially by the

456

conservative approximations, e.g., greater fuel

size and assembly reactivity, than will actually

he shipped. The criticality analysis conclu-

sively demonstrates substantia] subcriticality

under all of the conditions of temperature,

pressure, puncture, impact and exposure to

incidental water during anticipated maximum

potential damaging conditions of normal

transport.

3.0 Accident Evaluation

3.1 Free Drop Test

Height: 30 ft. measured by weighted string

hung from package.

Attitude of Long axis about 25° from vertical.

impact:

Impact surface Top edge of end which has the

of package: clamped end cap on the inner pack-

age. (From a series of drop tests

of an earlier prototype, it was

clearly established that the end

cap and cover edge was the most

vulnerable impact attitude. Previ-

ous drops also included flat on the

cover surface and flat on the bot-

tom.

Surface struck: 18” thick concrete pad. Pad not

damaged by test drop.

Contents: Two simulated fuel assemblies

each fabricated of 34” diameter

steel rods 157” long over a wood

core and welded to typical reactor

Results:

3.2 Puncture Test

457

fuel assembly hardware at both

ends. Dummy dimensions: 6” x 6”

x 174” long; weight 632 lbs. each.

Damage confined to impacted area.

End of outer box separated from

hody exposing inner container but

inner container still in position in

outer container. End cap of inner

container was still latched but gas-

keted surface not in contact with

shell flange along one edge of end

cap. Subsequent immersion would

have permitted water leakage into

inner container. End cap had to

be unlatched and pried off with a

tool in order to permit subsequent

removal of cover and fuel assem-

blies. Two-inch annulus remained

intact except in a 5” long area at

one end of the cover where result-

ing space averaged about one inch ;

an insignificant (1) portion of the

total container surface.

In view of the fact that the inner metal container

remained firmly within the outer packaging and,

since the inner package was properly designed to

remain subcritical with water inleakage such as that

which could result from a puncture, an actual punc-

ture test was not conducted.

3.3 Thermal Test

An actual thermal test was not conducted but the outer

packaging and cushioning material was assumed to

3.4

3.9

458

be completely consumed under thermal test condi-

tions. However, the remaining inner package is con-

structed of noncombustible material and the mini-

mum melting point of the fuel material is 4360°F.

The pressure relief valve permits escape of heated

air from the metal container. Therefore, this device

would prevent rupture of the container even if the

gasket did not melt to allow pressure relief. The

18-gauge steel container is internally braced with

lengthwise angle iron on both sides and bottom; the

cover is similarly braced. Previous tests of an 18-

gauge 55-gallon drum in a jet fuel fire which exceeded

an exposure temperature of 1475°F for more than 30

minutes produced no melting or distortion of the

outer surface. It is therefore concluded that a sim-

ilar thermal test at 1475°F for 30 minutes would

result in no damage to the Model RA-1 container nor

its contents more severe than the assumed condi-

tions considered in the criticality analysis presented

below.

Water Immersion Test

Since the package is designed to remain subcritical

assuming any degree or credible mode of water in-

leakage, it was unnecessary to subject the package

to an immersion test.

Analysis of Damaged Packages

From the above tests and assessments, the damaged

package is assumed to be as follows:

(a) All materials outside of the inner metal con-

tainer have been consumed by fire or otherwise

removed.

459

(b) Package contents (fuel assemblies) remain in-

side the metal container.

mately 2” thick around the fuel assemblies.

(d) The assemblies are of the most reactive type

which could be shipped under the authorization

requested herein, Le., 514” square cross-section,

64 rods of 0.580” diameter clad in 0.020” zir-

conium at 5% U-235 enrichment. The geometric

buckling of a reflected two-assembly container is

0.0188.

(e) The ko value of the assembly defined above is

taken at 1.54. This value is conservative since

the value used for design of fuel storage facili-

ties including storage pools is 1.4; maximum ko

values for presently planned fuel are approxi-

mately 1.3.

Based on the above factors, two cases were analyzed,

namely, the flooded and nonflooded inner package. If

the metal containers were flooded with water and

closely “stacked” in an infinitely-large, three-dimen-

sional array, the calculated ko value would be .954;

a 250 container array would have a calculated ko

value of 0.943. Assuming, more conservatively, the

annulus thickness had in some manner been reduced

over the entire surface of 250 containers from 2 to

1.5 inches, the array would still be subcritical.

In the second case, it was assumed that the metal

containers were submerged in, or sprayed with,

water but did not permit inleakage. Considering

differing degrees of water moderation that could

result from differing degrees of separation between

_ a

460

packages, a maximum k~ value of approximately 0.93

was calculated at the most reactive spacing of about

0.15 inches. The following table indicates values over

the range of analysis:

bebesen anskaieass (Calculated)

0 nm

15 93

50 78

1.00 69

1.50 53

These calculations considered effects of interspersed

moderation between the containers over a sufficiently

wide range of moderation to demonstrate suberiti-

eality under all credible conditions.

Caleulative methods incorporated standard reactor

design techniques utilizing Muft-type slowing down

calculations with heterogeneous corrections to ob-

tain two-group averaged epithermal cross-sections.

Cross-sections averaged over the thermal range were

obtained from solutions of the Wilkins equation com-

bined with a P-3 approximation to the Boltzman

equation. Accuracy of these techniques for systems

of this type has been well demonstrated by reactor

operating experience and other experimental data

for similar systems. These methods are described in

detail in Section 5 of the applications for license filed

under these dockets.

Thus, regardless of whether water were present as

the result of a spray, through immersion, or for any

other reason, the fuel assemblies would remain suh-

critical in any number. This same calculation demon-

461

strates subcriticality also for the “as-shipped” pack-

ages in which the wood and plastic material, either

wet or dry, contributes a certain degree of modera-

tion because of its hydrogen content.

In summary, therefore, an infinite number of pack-

ages, either in the damaged condition or undamaged

but in a water environment, would remain sub-

critical in any arrangement.

462

Supplementary Information

IV TRANSPORTATION:

HIGH LEVEL WASTES FROM THE PLANT

I. Basic Information

Waste material will be in several forms at the Vermont

Yankee plant. The largest source of waste material and

the only source of high level waste material will be the

spent fuel. Water treatment resins are used in various

plant process streams to purify and remove radioactive

constituents from them. The major portion of these ex-

hausted resins are low activity. Oiher smaller quantities of

miscellaneous radioactive wastes consisting of rags, floor

sweepings, worn out equipment and paper will also require

disposal. These items and their disposal will be discussed

individually.

Il. Spent Fuel

A. Description

The majority of all waste materials are contained within

the spent fuel assemblies and are shipped intact to the re-

processor. The physical dimensions of a spent fuel as-

sembly are identical to a new fuel assembly; about 1414

feet long x 5.4 inches square and weigh about 500' pounds.

Over the 30-year lifetime of the plant, about 2600 fuel as-

semblies will be used and require shipment to a reprocessor;

this is equivalent to approximately 3500 cubic feet and

1,300,000? pounds of waste material.

1 This is a typographical error in the original and should read “680”

instead of “500”. See FSAR (Ez. 1) §1.7, Table 1-7-1,