ML20205K881

Responds to Request for Addl Info Re Plan in Response to NRC Order CLI-88-07. Calculated Reactor Coolant Radioactivity Levels for Co-60,Cs-137 & Sr-90 Encl

ML20205K881
Person / Time
Site:	Seabrook
Issue date:	10/28/1988
From:	George Thomas PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
CON-#488-7402 CLI-88-07, CLI-88-7, NYN-88144, OL-1, NUDOCS 8811010382
Download: ML20205K881 (7)

Text

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George S. Thomas  ;

I NYN-88144 wee M so.newei.or hoovenon gggg October 28, 1988 Ph/w HampsNre Yankee Division United States Nuclear Reguintory Commission r Washington, DC 20555 Attention: Document Control Desk References (a) Facility Operating License NPF-56 Docket No. 50 443 (b) PSNH Letter (NYN 88142) dated October 20. 1988 ' Response to  !

CLI-88-07.' Edward A. Brown to USNRC r

Subject:

Request for Additional Information Regarding 'The Plan  ;

in Response to NRC Order CLI-88-07' Gentlemen:

! This letter and the attached analysis are in response to inquiries made l by the NRC Staff regarding 'The Plan In Response To CLI-88 07' (enclosure to Reference (b)) which describes and estimates the cost of activities that would j be necessary to return the Seabrook Unit 1 site to unrestricted use and to terminate the license after the facility has operated at up to 52 of rated  !

I power but not granted authorization to proceed to full power. i Table 1 in Section 7.5 of the Plan indicates that isotopes of Cobalt-60, d

Cesium-137 and Strontium-90 were not detected in the reactor coolant of the j three plants surveyed. each of which had limited operation. During low power j testing. Seabrook will experience less than one Effective Full Power Hour of j operation and, therefore, similar results would be expected. The analysis l shown on Attachment 1, however, assumes that concentration of these isotopes  ;

are exactly at their lowest levels of detectability in order to determine the  !

maximum hypothetical surface contamine m ,.: that might result. It was also '

conservatively assumed that there was no removal of these isotopes by filtration and that plateout of these isota.es was all concentrated in the  ;

steam generators. As shown, even under these assumptions, the resultant  ;

surface contamination levels would still be well within the Regulatory Guide i 1.86 criteria for unrestricted release. [

l It should be noted that even if one assumes that the surface  !

contamination levels determined by the Attachment 1 analysis were above the  !

Regulatory Guide 1.86 criteria, this contamination could be effectively  !

eliminated by a combination of acid and potable water flushes. The Plan, as i submitted, contemplates four potable water flushes. The cost of substitution i of an acid (42) flush for one of these could easily be accommodated by a i combination of the decontamination cost estimate and the contingency and would  !

not change the overall $21.1 million cost estimate to implement the Plan.  !

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8s11010382 ADOCK g % 43 pDR PDC P.O. Box 300. Seabrook, NH 03874 . Telephone (603) 474 9574 /

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, Unite,d Statos Nucloor Regulatory Consnission Octobcr 28, 1980 Attention: Documsnt Control Dosk Pago 2

Vith regards to the subject of shipment and disposal of fuel, a number of l' options were considered. One of these options involves selling the slightly irradiated fuel to another utility. In view of the very limited operation of the fuel during the low power testing program, this was considered feasible.

4 However, this option was not chosen for the estimate prescribed in the Plan.

Instead, since the decomissioning rule provides that '[djecomissioning d

activities [are not to) include the removal and disposal of spent fuel...".

4 NHY conservatively assumed overseas reprocessing becaus in this option, the fuel would remain onsite for the longest period of time If you have any further questions concerning this matter, please contact 1 Mr. Robert Sweeney at (301) 656-6100.

! Very truly yours, r$ /

j George S. Thomas l Enclosures cc: Atomic Safety and Licensing Board Service List

, Mr. Villiam T. Russell

Regional Administrator i United States Nuclear Regulatory comission Region I 475 Allendale Road King of Prussia, PA 19406 Hr. Victor Nerses, Project Manager

-l Project Directorate I-3 Division of Reactor Projects United States Nuclear Regulatory Coumission Vashington, DC 20555 1

i Mr. David G. Ruscitto NRC Senior Resident Inspector P.O. Box 1149 Seabrook, NH 03874 I

i I

Table 2 of Section 3 of The Plan (cash flow) shows that the costs I

associated with storing fuel onsite after completion of decontamination and

< removal of the reactor vessel and associated equipment is approximately I

$110.000 per month.

Attcchment to NYN-88144 ATTACRMENT 1 [

t Calculated Reactor Coolant Radioactivity Levels For Co-60. Cs-137 and Sr-90 1

Background

Some typical radioisotopes found in operating nuclear power plants f.re Cobalt-60, Cesium-137 and Strontium-90. These nuclides (Co-60, Cs-137) were not 4

detected in the primary coolant samples of three recent vintage pressurised water reactors (plants A, B C) with low operating history of less than or equal  ;

to ten effective full power hours (Ref: "The Plan In Response To NRC Order i CLI-88-07", page 7-25). The following analysis, however, considers the effects  !

of assuming concentrations of these isotopes (Co-60, Cs-137) ar the lowest level }

of detectability (LLD) with a derived concentration of Sr-90 bared on the Cs-137 value. (

Assumptions j The following conservative assumptions were used in the calculation of nuclide  !

activity and deposition levels. l i

1. Both Co-60 and Cs-137 exist in the reactor coolant system at the highest  !

reported LLD value with no removal by plant filtration systems.

I i 2. Since Sr-90 is non a photon emmitter and thus is not detectable using ,

gamma scintillation methods, the St-90 concentration is calculated using a l St-90/cs-137 fission product ratio applied to the Cs-137 LLD. This level is i assumed to exist in the reactor coolant system with no removal by plant i

filtration systems.

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3. One hundred percent of the calculated Co-60, Cs-137 and Sr-90 inventory is l plated out within the surface area of the steam generators.  :

Data r

1. LLD for Co-60, Cs-137 for plants A, B, C r

LLD (uCi/g)

Nuclide Plant A Plant B Plant C L Co-60 1.2E-07 7.5E-09 1.7E-07 l Cs-137 1.2E-07 1.3E-08 1.8E-07

2. Plant data (Seabrook Station Final Safety Analysis Report Tables 5.4-3, 11.1-1, 11.1-3) l

a. Sr-90/Cs-137 = 5.6E-04 (Basist Reactor Coolant Concentration, .121 i clad defect) i

b. Reactor coolant mass = 2.29E+08g [

c. Steam generator (4) tube surface area = 2.04E+08cm2

. i

, Attcchasnt to NYN- 88144 Calculations

1. Activity (uCl)

A(Sr-90) = s5.6E-04) (2.0E-07uct/g) (2.29E+08g) = 2.56E-02uci A(Cs-137) = (2.0E-07uci/g) (2.29E+08g) = 4.6E+01uci A(Co-60) = (2.0E-07uci/s) (2.29E+08g) = 4.6E+0lu01

2. Steam Generator Plateout (dpm/100cm2) at 100% plateout Sr-90 (2.56E-02uci) (2.22E+0cdpm/uci) (100)/(2.40E+08cm2) . 2.8E-02 Cs-137 (4.6E+0luci) (2.22E+06dpm/uci) (100)/(2.04E+08e.2) = 50 Co-60 (4.6E+0luci) (2.22E+06dpm/uci) (100)/(2.04E+08cm2) = 50 TOTAL 100dpm/100cm2 Results The total calculated contamination levels of Co-60, Cs-137 and Sr-90 on steam generator surfaces would be less than 100 dpm/100 square centimeters. This is only 10% of the USNRC Regulatory Guide 1.86 release limit.

A review of nuclide deposition data found in EPRI-NP-5900 was performed along with information in EPRI-NP-2968 which was a basis document for NP-5900 data.

Figure 2.1 from NP-2968 indicates that the slope of the curve associated with steam generator Co-60 surface activity changes significantly as a fun < tion of the amount of power operation (EFPY). The slope of the curve is non-linear with activity levels very close to zero for very low effective full power histories (less than .5 ETPY). For this reason, extrapolation of the curve to a level close to one Effective Full Power Hour (EFPH/ vou'.d be inappropriate using the given EPRI data. For purposes of the above cale dation, it was concluded that the use of plant empirical data is more represertative when considering very low EFPH which would result in minimal wear associrced with operation of plant valves as in the case of the Seabrook low power test program and the operating history froa plants A, B and C.

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Diane Currap. Esquire  :

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Atomic Safety and Licensing Appeal Panel Harmon & Weiss  !

U.S. Nuclear Regulatory Connission Suite 430  !

East Vest Towers Building 2001 S Street, NW f 4350 East West Highway Vashington, DC 20009 i Bethesda, KD 20814 {

Stephen E. Herri11. Esquira  !

Administrative Judge Sheldon J. Wolfe, Attorney General .

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• Atomic Safety and Licensing 4350 East West Highway Appeals Board Panel l Bethesda, MD 20814 U.S. Nuclear Regulatory Commission

• Vashington, DC 20555 .

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• Senator Gordon J. Humphrey Mr. Peter S. Matthews United States Senate Mayor Vashington, DC 20510 City Hall Attention: Tom Burack Newburyport. HA 01950  !

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• Senator Gordon J. Humphrey Mr. Villiam S. Lord ,

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Backus. Heyer & Solomon Newburyport. HA 01950 [

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City Hall Rockville. HD 20852 I 126 Daniel btreet Portsmouth NM 03501 i

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