ML20090L585
| ML20090L585 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 02/13/1976 |
| From: | Mayer L NORTHERN STATES POWER CO. |
| To: | Stello V Office of Nuclear Reactor Regulation |
| References | |
| 1508, NUDOCS 9102120440 | |
| Download: ML20090L585 (9) | |
Text
_. _
NSF NOMTHEMN STATES POWER COMPANY MIN N E A POL,0. MIN N E S OTA 5 5 tot
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Mr Victor Stello, Director
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D'O',7 Il U S Nuclear Regulatory Coctnission L,
Division of Operating Reactors q, 2 8
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N Washington, DC 20555
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'l Dear Mr Stellos
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HONTICEL1h NUCLEAR GE!$ RATING PIANT Docket No. 50-263 License No. DPR-22 Of f-Site Shipment of Spent Fuel on January 22, 1976 we transmitted to you a report entitled "An Analysis and Safety Evaluation of Spent Fuel Shipping Cask Handling at the Monticello Nuclear Generating Plant" dated January 13, 1976.
Since then we have had several telephone conversa-tions concerning this report with Mr R P Snaider of your staff. Mr Snaider had several questions concerning the e.onsequences of a cask drop down the Reactor Building equipment hatch.
One area of concern was the potential for the cask to impinge upon the suppression pool if the cask should drop in other than a vertical attitude. C n ideration was given to this situation in our review of cask handling operations and it was determined not to be a relevant consideration for several reasons. The cask travel path, shown on Figure 6-1 of our January 13 report, was selected such that the cask would be moving away fecun the suppression pool while it was still in the hatchway opening.
If the cask should drop while in the hatchway, it would then land on the dividing wall or to the west of it, away fro:n the suppression pool.
It is not considered credible for the cask to fall in other than a vertical attitude unless it is forced into a tilted position before the drop occurs.
This could occur only if the cask is moved along travel path A-B before it has cleared the hatchway opening at elevation 1027'-8".
This event will be precluded by ad:ninistrative l
controls and rehearsals of fuel shipping operations before the cask is lif ted onto l
l the 1027'-8" level.
l Mr Snaider also asked for the offsite radiological consequences that could be expected should a loaded cask be dropped down the equipment hatch. This situation was analyzed and the potential consequences were found to be significantly lower than the limits specified in 10CFR100. The acalysis that was perfomed is suttnarized in the attached report.
I 1508 9102120440 760213 PDR ADOCK 05000263 P
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l NORTHERN OTATED POWER COMPANY
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j Mr Victor Stello Page 2 l
February 13, 1976 r
I We are anxious for a resolution on this matter and suggest that a meeting with members of your staff be arranged at your earliest convenience should you have any l
further questions in regard to our fuel shipping plans.
1 Yours very truly, i
WC.
j L 0 Mayer, PE Nbnager of Nuclear Support Services
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1 I
POTENTIAL RADIOLOGICAL CONSEQUENCES OF CASK DROP DOWN THE REACTOR BUILDING EOlti? MENT HATCH 4
The radiological consequences at two offsite locations (the site boundary:
500 meters, and the low population zone:
2 miles) were determined for the drop of a t
two element shipping cask down the reactor building hatchway.
A.
Assumptions
]
The following is a summary of the assumptions that were made in evaluating the off-site doses i
1.
All of the assumptions in Regulatory Guide 1.25 were used with the following exception:
Pasquill diffusion data given in Regulatory Guide 1.3 and 1.25 were used because equivalent site meteorological data was not available from the Monticello FSAR.
1 2.
The accident is assumed to occur with the containment isolated at a negative containment pressure of 0.25" H 0.
As a result, 2
all releases to the environment are assumed to exit via the Standby Gas Treatment System (SGTS) with a subsequent elevated releare from the offgas stack which has a release point 125 meters above the surrounding grade.
3.
Each train of the SGTS is capable of replacing the containment atmosphere at the rate of one air change per day (Reference 2).
Only one train was assumed to 'oc available at the time of the accident.
4.
The gases released from the damaged rods are released to the containment immediately with no holdup in the fuel rods.
5.
Radioactive inventory is proportional to fuci power level.
6.
No credit is taken for decay of the gaseous radioactivity while in transit from the stack to the recipient.
7.
It was assumed that a two element shipping cask is dropped down the reactor building hatchway and damaged such that all of the 98 rods within the cask release their gaseous inventory.
8.
Plant release via SGIS, without mixing with the containment atmosphere, is made over a two-hour period (Reference 1) 9.
The two bundles considered are of maximum burnup with a radial peaking f actor of 1.5 for all 98 rods.
10.
Fumigation conditions exist for the first 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, followed by 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of normal abmospheric dispersion.
(Reference 1) i l
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B.
Radiological Model The dose received at an offsite location due to the release of the gaseous radio-active inventory of a damaged fuel rod into the containment must be estimated.
The dose was calm lated as the product of the following series of terms:
1.
(Sj) average scurce strength of isotope j in each rod at
=
the time of plant shutdown (C1).
2.
(P )
e i
peaking factor for the rod being considered (rod i).
3.
(GF))
= fraction of the total rod inventory of isotope j in the plenum and available for release.
4.
(DFj) fraction of the original activity of isotope j which a
remains after decay due to storage in the fuel pool.
This term also takes into account any production of isotope j due to the decay of a parent during this time.
5.
(1-N)) m fraction of isotope j vhich is not trapped in the SGTS filters and is allowed to escape via the offges stack.
(Reference 1) 6a.
(RF ) c k
fraction of the total radioactive containment inventory which is released to the environment during the time period under consideration.
6b.
(BR ) a k
breathing rate (for inhalation dose only) for the standard man during the time period under consideration.
Breathing rates do not affect the Gamma and Beta (skin) doset, and l
this factor is therefore taken as 1 for the calculation of E
skin doses.
6c.
(X/Q }z e Pasquill diffusion coefficient at locagion (z) which is k
applicable during time period k (sec/m ).
Since (RF )
( ER ), and (X/Q )z are all functions of the time period (k) under k
k k
consideration, the sum of their products will give the required integrated term over the total period of the accident, i.e.,
k = (RF ) (BR ) (X/Q )z, then the required term to be used in Let C k
k k
2C, where the calculation of the dose received at location z is k
Ck C1+C2........+C.
=
k k
(DC))
a dose conversion factor for the jth isotope. This converts 7.
the integrated concentration of activity with respect to time into a dose in rems (rem /ci for inhalation or 3
rem-m /ci-sec).
Therefore, the dose (in rems) received at location z due to isotope j, frca rod i, is (S ) (P() (GFj) (DF ) (1-N ) (DC))
C j
j j
h (1)
s.
To find the total dose received from one rod we sum over all j isotopes involved.
1 To find the total dose received from all rods, a sum over all i rods is performed.
l The overall dose, D in rems at offsite location z ist D=
(S)) (P ) (GFj) (DF)) (1-Nj) (DC))
Ch (2) t j
i j
k C.
Input Datn 1.
Isotopic Source Terms (S,) (Reference 3)
Isotope Ci pes Average Rod I-131 5297 1-132 728 1-133 139 i
I-134 2.5E-5 I-135 171 Xe-131m 18.8 Xe-133 3655 Xe-133m 65.9 i
Xe-135 1075 Xe-135m 1.01 l
Kr-83m
.12 Kr-85 248 Kr-85m 3.45 Kr-87 2.5E-4 Kr-88
.71 2.
Pasquill Diffusion Coefficients (Refererce 1)
For 125 meter release height; i
. (X/0)
Time Period Af ter Accident 500 Meters 2 Miles i
0 - 1/2 hr.
1.6 x 10-4 3.2 x 10-5 1/2 - 8 hrs.
1.4 x 10-5 6.3 x 10-6 8 - 24 hrs.
2.0 x 10-5 J.2 x 10-6 1 - 4 days 3.0 x 10-6 8.1 x 10-7 i
4 D.
Radiological Analysis Results The dose consequences for the equipment hatch drop case are based on equation (2) of Section B.
The tabulated results as a function of shipping date for a cask drop in which all of the 98 rods contained in two fuel bundles are assumed f ailed is presented in the follwing Table.
Regulations in 10CFR100 require that offsite doses be limited for accident situations tot 1.
25 rems whole body 2.
300 rema thyroid Based on the results, the conservatively calculated doses for this postulated accident fall well bel w 10CFR100 limits.
Shipments of spent fuel in two element shipping casks may thet efore be made at any time af ter January,1976, with negligible impact to public health and safety should this postulated accident occur.
~
DOSE RESULTS FOR EQUIPMENT llATCil DROP CASE Done in Rems Date of Shipment 500m 2mi Thyroid Whole Body
- 7. tr oi d Whole Body January 1976 88 x 10*3
.028 21 x 10-3 7.1 x 10-3 April 1976 38 x 10-6
.027 97 x 10-6 7,o x 10-3 July 1976 17 x 10-9
.027 4 x 10-9 6.9 x 10-3 October 1976 nil
.026 nil 6.8 x 10-3 January 1977 nil
.026 nil 6.7 x 10-3
o REFERENCES 1.
USNRC Regulatory Guide 1.25, Assumptions Used for Evaluatine the Potential Radiological Consequences of a Fuel llandling Accident in the Fuel llandling and Stor8Re Facility for BoilinR and Pressurized Water Reactois, March, 1972.
2.
Monticello FSAR, Chapter XIV, Safety.'.un1.ysis.
3.
G.E. Standard Safety Analysis Report (CESSAR), Chapter 15, 1
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fit 0M: NORTHERN STATES POWER CO 0*T80'00CV"EN7 HR V ST M 0 HINNEAPOLIS, HINN.
L 0 MAYER 2 17 76
%atitn ONOTOHt2ED PAOP INPUT FORM NUMBE R OF COPIEF RECEIVED kNC LASSIFIE D nlOIN AL 39 Cory OtscntPT80N E NC LOSU RE LTR, RE THEIR l-22 76 SUBMITTAL "AN ANALYSIS REPORT CONCERNING THE CONSEQUENCES OF A CASK DROP DOWN THE REACTOR BUILDING EQUIP.
6 SAFETY OF SPENT FUEL.....TRANS THl:
HENT HATCH......
I FOLLOWING....
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PLAlff NAME: L MONTICELLO i
SAFETY ~~
FOH ACTION /INFORMAllON ENVIRO
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Aff1'GiMD AD :
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ASSIGNED AD :
Bliftfdl CIIIEP :
&Lgmn A BRANCiiCHIEF v
PROJECT T1XRAGER:
PRO' JECT MANAGER 't V
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_sCHROEDER BENAROYA BAIJ.AR_D SPANGLER ih _0 ELD
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COSSICK & STAFF ENGilEERING IPPOLITO L_ MIEc MAccMY SnF. TECH L__. CASE KNIGHT OPERATING REACTORS GAMIILL
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__IIARLESS PAWLI_C,KI ll0IRAh OPERATING TECll
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'~ PROJECT MANAGEMENT REACTOR SAFETY
& _ EISEN}Lttr SITE ANALYSIS BOYD ROSS N
VOLIJfER SilAO P. COLLINS NOVAK
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BAER BUNCH HOUSTON ROSZTOCZY
& _SCIMENCER J. COLLINS PETERSON Cl!ECK
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GRIMES KREGER MELTZ
~ AT & I SITFJAFE_IY_& EINIRC HELTEMES SKOVHOLT
,_SALTZMAN ANALYSIS RITrBERG DENTON & HULT.ER EXTERNAL DIS ~l HIBUTION CONTROL NUMBER y 3pg, iiihdaiwa5,n:;
NATLLI.Ali BROOKilAVDi NATL LA5-REG. V-IE ULRIKSON(ORNL)
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