Text

Responds to NRC Comments on Environ Issues Re Bg&E License Application for Facility Isfsi.Encl Proprietary Info Withheld (Ref 10CFR2.790)
	ML20058F018
Person / Time
Site:	Calvert Cliffs
Issue date:	11/01/1990
From:	Denton R; BALTIMORE GAS & ELECTRIC CO.
To:	; NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
Shared Package
ML19310C824	List: ML20058F018;
References
	NUDOCS 9011080061
	Download: ML20058F018 (55)
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ES ALTIMORE GAS AND ELECTRIC CHARLES CENTER e P.O. BOX 1475

BALTIMORE, MARYLANO 21203 1475 R E.oENTON GENERAL MANAGER CAlvtR1 CLIFFS November 1,1990 j

U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:

Director, Office of Nuclear hiaterial Safety and Safeguards E

SUBJECT:

Calvert Cliffs Nuclear Power Plant Independent Spent buct Storage Installation; Docket No. 72-8 (50-317/318)

Response to NRC's comments on Environmental issues Regarding BG&E's K

License Applicaticn for Calvert Cliffs Independent Spent Fuel Storage Installation (ISFSD

REFERENCES:

(a) Letter from. Mr. F. C. Sturz (NRC) to Mr. G. C. Creel (BG&E),

dated July 23, 1990, Comments on Safety Analysis Report and Environmental Report E

(b) Letter from Mr.

G.. C. Creel (BG&E) to Director, Division of Z

industrial ' and. Medical Nuclear Safety Office of Nuclear Material i

Safety and Safeguards (NRC), dated December 21, 1989, Calvert Cliffs ISFSI Application j

H Gentlemen:

i This is in response to the Nuclear Regulatory Commission's (NRC) comments (Reference a) on Baltimore Gas and Electric Companyi (UG&E) License Application for

{

Calvert Cliffs Independent Spent Fuel Storage Installa: ion (Reference b). Transmitted ar Attachment A to this letter is our response, which addresses your comments on the environmental and radiation protection portions of our license application.

Attachment B is a proprietary affidavit pursuant to 10 CFR 2.790, nnd Attachment C contains proprietacy documents in suppo t of this response. We expec

o submit responses for the remainder of the Safe > Analysis Report comments by

...e end of l

6 December 1990.

Our present plans call for the first phase of construction ta begin as early as December 1990. In order to: meet this schedule, we respectfully request that the NRC

'fj/

expedite review and approval of the Environmental Report based on the additional information contained herein.

90.11000061 901101 PDR ADOCK 05000317

,,Y, PDC

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Office of Nuclear Material Safety and Safeguards November 1,1990 Page 2 Additionally, the Safety Analysis Report and the Environmental Report will be revised as a result of these responses. Upon completion of your review, we will submit the revised documentation for your records.

Should you have any furthei questions regarding this matter, we will be pleased to discuss them with you.

Very truly yours, lr a

an 1

i GCC/GT/dtm Attachments:

As Stated cu (Without Attachment C)

D. A.11 rune, Esquire J. E.

Silberg, Esquire R. A.Capra, NRC W. T. Russell, NRC D. G. Mcdonald, Jr., NRC L. E. Nicholson, NRC F. C. Sturz, NRC R.1. McLean, DNR With All Attachments F. C. Sturz, NRC (5 Copies)

S. M. Mirsky, SAIC (1 Copy) i i

i I

Office of Nuclear Material Safety and Safeguards November 1,1990 Page 3 bec: (With Attachment A & 11)

J. A. Tiernan/C. J. Franklin, Jr.

M. J. Micrnicki R. F. Ash E. I. Bauereis/M. C. Gavrilas W. R. Corcoran C. H. Cruse /P. E. Katz R. C. DeYoung R. M. Douglass R. P. Ileibel/T. N.

Pritchett C. P. Johnson C. C. Lawrence, Ill/A. R. Thornton R. B. Pond, Jr./S. R. Buxbaum L. B. Russell /J. R. Lemons G. L. Adams (2)

A. B. Anuje J. E. Baum G. L. Bell J. J. Connolly R. E. Denton G. L. Detter G. J. Falibota L. D. Graber L. S. Larragoite B. S. Montgomery P. A. Pieringer J. II. Walter M. J. Warren R. II. Dufrense M. C. Key M. L. Stone W. J. McConaghy A. M. Segrest P. A. File (With All Attachments)

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RESPONSE TO NRC COMMENTS ON TEE CALVERT CLIFFS NUCLEAR POWER PLRNT ISF8I ER 1:

QUESTION:

ER-1 Please provide copies of the following BG&E drawings.

"NUHOMS'-24P ISFSI Horizontal Storage Modules Preliminary Site Plan," Drawing No. BGE-01-2000.

"NUHOMS*-24P ISFSI Horizontal Storage Modules General Layout,"

Drawing No. BGE-01-2001, i

"NUHOMS'-24P ISFSI Horizontal Storage Modules," Drawing No.

BGE-01-2002.

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RESPONSE

The requested drawings have been superseded by the following series of drawings which are attached:

" Independent Spent Fuel Storage Faciiity - Site Plan," Drawing l

No. 84-075-E.

" Independent Spent Fuel Storage Facility. HSM Concrete Plans," Drawing No. 84-080-E.

" Independent Spent Fuel Storage Facility HSM Concrete Sections," Drawing No. 84-081-E.

" Independent-Spent Fuel - Storage Facility HSM Concrete Sections and Details," Drawing No. 84-082-E.

" Independent Spent Fuel Storage Facility HSM Reinforcing Plans," Drawing No. 84-085-E.

" Independent Spent Fuel Storage Facility HSM - Reinforcing Sections," Drawing No..84-086-E.

" Independent: Spent Fuel Storage Facility HSM - Reinforcing _

Sections," Drawing No. 84-087-E.

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" Independent Spent Fuel Storage Facility Reinforcing Bill of Materials - HSM," Drawing'No. 84-088-E.

" Independent-Spent' Fuel Storage Facility HSM - Miscellaneous Steel Plans, Sections and Details," Drawing No 84-091-E.

" Independent Spent Fuel Storage Facility HSM - Miscellaneous Steel ~ Plans, Sections and Details," Drawing No. 84-092-E.

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- BGE001.0024.03

RESPON8E TO NRC COMMENTS ON THE CALVERT CLIFFS NUCLEAR POWER PLANT ISF8I ER (Continued)

QUESTION:

ER-1 (Continued)

" Independent Spent Fuel Storage Facility HSM - Miscellaneous Steel Plans, Sections and Details," Drawing No. 84-093-E.

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BGE001.0024.03

i RESPONSE TO NRC COMMENTS ON TER CALVERT CLIFFS NUCLFAR POWER PLRNT ISFSI Elt QUESTION:

ER-2 Para. 3.1, p. 3.1.1 Clarify what is meant by " nominal doses."

RESPONSE

1 The term nominal doses refers to the area averaged dose rate on the HSM surfaces.

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RESPONSE TO NRC CCitMENTS ON THE CALVERT CLIFF 8 NUCLEAR F3WER PLANT ISF8I ER QUESTION:

ER-3 a.)

Para.

4.4, p.

4.4-1 What is the bases for assuming a dose rate of 0.5 mrem /hr?

b.)

Provide graphs of dose rate vs. distance (close in, &l00 ft.) from the face and side of one row of modules.

c.)

What ALARA considerations have been given to HSM loading patterns in relation to the timing of construction for the next row of modules and to the potential reduction of construction worker exposure? What incremental number of modules will be constructed at one time, one 2x6 array, one 2-2x6 row, or several rows?

d.)

What is the maximum annual occupational exposure, for construction, expected in any one year?

RESPONSE

a.)

A dose rate of 0.5 mr/hr was assumed in the submittal of the ER for a bounding dose rate during construction of HSM arrays after fuel has been loaded into the ISFSI.

Detailed calculations were completed after the ER had been submitted which show that the maximum dose rate to any worker during construction is 0.11 mrem /hr.

This dose rate is based on the assumption that each of the HSM's in 3 1/2 phases contains 24 de'aign basis fuel assemblies, the HSMs facing the construction area have not yet been filled and Phase 5 is under construction.

No credit is taken for source decay alter storage.

The 0.11 nrem/hr value includes the effer,ts of a revised HSM door design which reduces area dose rates below those predicted in the ER.

The SAR and ER will be revised to incorporate the new door design and the resulting changes i r. ISFSI dose rate information.

b.)

The attached figures contain the plots of dose rate vs.

distance from the side and face respectively of one row of modules (a single construction phase containing 24 HSMs) and five. rows of modules (all five phases containing 120 modules) when the modules are filled entirely with design basis fuel assemblies. These figures include the effect of an improved HSM door shield design described above.

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c.)

Initially, at least one HSM phase will be constructed,_

l consisting of two 2x6 HSM arrays placed end to end.

l BGE001.0024.03 l

RESPONSE TO NRC COMMENTS ON THE CALVERT CLIFFS NUCLEAR POWER PLRNT Z8F8I ER (Continued)

I l'

QUESTION:

ER-3 (Continued)

During construction of the subsequent phases, the HSM row facing the construction activities will remain empty.

This empty HSM row will provide shielding during construction of the subsequent phase. This strategy will l

eliminate the direct component of the radiation exposure i

during construction, thus minimizing construction l

exposures.

However, based on actual dose rate l

measurements, additional HSM construction and loading l

will be evaluated to ascertain the need to maintain an empty HSM row for shielding purposes, d.)

Assuming that 24 modules will be built each year and 36,000 manhours are required to build these modules, the maximum annual occupational exposure for construction is estimated to be 3,960 person-mrem. This is based on the following calculation:

0.11 mrem /hr *36000 hr = 3,960 person-mrem l

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BGE001.0024.03

RESPONSE TO NRC COIOtENTS ON TEE CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI ER (Continued)

QUESTION:

ER-3 (Continued)

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BGE001.0024.03

RESPONSE TO NRC CONNENTS ON THE CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI ER (Continued)

QUESTION:

ER-3 (Continued)

Cakert Cliffs !$FSI Dose Rate Study 0

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RESPONSE TO NRC COMMENTS ON THE CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI ER QUESTION:

ER-4 a.)

Para.

5.2.1, p.

5.2-1 (also SAR para 7. 4.1, p.7.4-1).

Due to the phased nature of the prr,j ect,

estimated occupational exposures for HSM loading operations are inexact. How were radiological exposures from previously loaded HSMs factored into the estimates for HSM loading operations in SAR Table 7.4-17 b.)

What is the estimated occupational exposure to other personnel not performing a particular task listed in the SAR Table 7.4-1 but who may be in the vicinity providing support-functions?

c.)

How many DSCs/HSMs are estimated to be loaded each year?

Provide an estimated upper bound for annual ISFSI occupational exposure.

RESPONSE

a.)

SAR Table 7.4-1 does not include the dose contribution from previously loaded HSMs since they are quite small compared to the dose received from the HSM and cask being loaded.

This is due to the combination of relatively

.high radiation fields near the transfer cask, and the relative proximity to the cask which is necessary to perform the loading tasks.

In order to estimate the dose due to previously loaded modules, we will use a general area dose rate due to previously loaded HSMs of 2.0 mrem /hr.

This value is a conservative estimate as can be seen by extrapolating the graph of dose rate vs. north-south distance provided in response ER-3.

The dose rate perpendicular to the front i

' face of a phase is less than 1 mrem /hr.

This value should-be doubled, however, to account for the faces of 2

two loaded' phases being visible.

The general area dose 1This is true everywhere except very near.to previously loaded modules, where the area-averaged dose rate is calculatad co be 8

' mrem /hr.

Since the HSM loading activities are around the casx, ram, and HSM to be loaded, and since good ALARA practices would include a low dose waiting area, the areas within a few feet of the already loaded HSMs were not considered here.

2Note that this effectively doubles the skyshine radiation contribution which is a significant contributor.

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~ BGE001,0024.03

RESPON8E TO NRC COMMENTS ON THE CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI ER (Continued)

QUESTION:

ER-4 (Continued) in an " alley" between two filled phases will be less than 2.0 mrem /hr.

Eight effective man-hours at the HSM site were estimated for one HSM load (refer to SAR Table 7.4-1).

The personnel exposure received during this time from previously loaded HSMs can therefore be estimated as 16 person-mrem.

By comparison, this represents 16/1455, or about 1% uf the expected loading operation exposure of 1455 person-mrem (with the revised HSM door geometry),

b.)

The personnel required to perform loading operations are accounted for in the SAR Table 7.4-1 including all support personnel who will be inside the controlled area boundary of the ISFSI. The only other support personnel required are the ISFSI security guards who will be i

stationed at the ISFSI security gate. The maximum dose rate received by each security guard will be 0.58 mrem /hr.

Assuming that the guards are present at the ISFSI security gate for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months whenever fuel is loaded, the total occupational dose received by these security guards will be 5.8 mrem per person.

Assuming a maximum of three security guards present at any time during the loading operatior., this will result in a total maximum dose of 17.4 persca-mrem.

c.)

A maximum of sixteen DSCs/HSMs are planned to be loaded each year. The upper bound for annual ISFSI' occupational exposure is estimated as follows:

1.

The maximum occupational exposure during one DSC/HSM load = 1455 person-mrem. The 1455 mrem /hr value includes the reduced HSM door dose due to revised HSM door design.

2.

The maximum occupational exposure during one DSC/HSM load due to previously loaded HSMs 16

=

person-mrem.

3.

The maximum occupational exposure during one DSC/HSM load to the additional support personnel (ISFSI security guards) = 17.4 person-mrem.

BGE001.0024.03 l

RESPONSE TO NRC COMMENTS ON THE CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI RR (Continued)

QUESTION:

ER-4 (Continued)

Based on these results, the maximum occupational exposure due to one DSC/HSM load is 1455 + 16.0 + 17.4 = 1488.4 person-mrem.

This will give a maximum annual ISFSI occupational exposure during DSC/HSM loading operations of 1488.4

16 = 23,814 person-mrem.

Some exposure is also received during the daily visual inspection of HSM air vents.

The eetimated annual occupational exposure received during the daily visual inspection of HSM air vents is 46 person-mrem.

The maximum annual ISFSI occupational exposure received during all ISFSI operations = 23,860 person-mrem.

This value does not include contribution due to reactor operation or other radioactive sources.

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RESPON8E TO NRC COMMENTS ON THE CALVERT CLIFF 8 NUCLEAR POWER PLANT ISFSI ER QUESTION:

ER-5 Para. 5.2.1, p. 5.2-1 (also SAR Para. 7.4.1, p. 7.4-1).

What is the estimated incremental increase in occupational dose to all site workers?

Provide a tabulation of the number of workers by location (or function), the estimated dose rate at that location, and the annual collective occupational exposure i

I at the location.

RESPONSE

l The tabulation is provided below.

The dose rate estimates are based on recent calculations performed by Pacific Nuclear Fuel Services, Inc.

They assume all five planned HSM phases are filled with design basis fuel.

No credit is taken for shielding by structures.

A 50 hour5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months / week occupancy factor has been assumed resulting in a total of 50*52=2600 work hours per year.

These dose rates include consideration for an improved HSM door shield design which will reduce area dose rates below those credicted in the Environmental Report.

Estimated Collective Location Number of Dose Rate Exposure Workers (mrem /hr)

(man-rem /yr)

OTF 240 1.2e-03 0.75 NEF 280 2.2e-03 1.60 MPF

-240 2.2e-03 1.37 l

NOF 470 4.7e-03 5.74 Inside 1200

2. 2 e-03*

6.86 the Protected Area of the Plant sum--->

16.32

NEF dose rate used as conservative estimate for the plant protected area.

BGE001.0024.03

e RESPONSB TO NRC COMMENTS ON THE ChLVERT CLIFF 8 NUCLEAR POWER PLANT ISF8I ER QUESTION:

ER-6 a.)

para. 5.2.1, p. 5.2-1 ( also SAR Para. 7. 4.1, p. 7.4-1).

What is the maximum total annual ISFSI occupational exposure, from construction, operation, and surveillance?

b.)

How does this compara to the anr.ual occupational exposure from existing reactor operations?

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RESPONSE

a.)

The maximum annual ISFSI occupational exposure from construction 3,960 person-mrem (see response to

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question ER-3-d).

The maximum annual ISFSI occupational exposure from operation and surveillance 23,860 person-mrem (see

=

response to question ER-4-c)

This results in the maximum total annual ISFSI occupational exposure during construction, operation and surveillance of 23,860 + 3,960 27,820 person-mrem, 1

=

during a year in which 24 HSMs are built and 16 canisters are loaded into HSMs.

b.)

In 1987, 1988 and 1989,_ Calvert Cliffs had a total annual occupational radiation exposure from existing reactor operations of 413 rem, 291 rem, and 346 rem respectively.

The 3-year average from reactor operations is 350 person-rem.

The projected mcximum annual ISFSI occupational exposure from operation (loading 16 HSMS) and routine surveillance is 23.86 person-rem.-

The maximum dose due to construction of 24 HSMs in a year is 3.96 person-rem.

The total resultant maximum annual exposure from the ISFSI is 27.8 person-rem.

This represents a small fraction of the exposures typically experienced during existing reactor operations.

l BGE001.0024.03

RESPONSE To NRC COMMENTS ON THE CALVERT CLIFF 8 NUCLEAR POWER PLANT ISFSI ER QUESTION:

ER-7 Para.

5.2.1, P.

5.2-1 (also SAR para 7.4.2, p. 7.4-1).

(see also SAR Questions 7.0-4).

The dose rate at the ISFSI restricted area fence (no closer than 50 ft.),

during storage only, approaches the 2 mrem /hr limit of 10 CFR Part 20.105 (b) (2).

Describe how BG&E will comply with the limits in 10 CFR 20.105 as well as to the 100 mrem /yr limit in the proposed revision to 10 CFR Part 20.

RESPONSE

10CFR20.3 defines unrestricted areas as any areas where access is not controlled by the licensee for purposes of radiation protection.

Pursuant to 10CFR20.105, three criteria must be satisfied for radiation in unrestricted areas.

The limits are:

500 mrem in any one year, no occupancy factor specified, 2 mrem in any one hour, 100 mrem in any seven days, 100% occupancy factor.

For the Calvert Cliffs ISFSI, the unrestricted access boundary is the outer ISFSI security fence.

For this design, the 10CFR20.105 bounding limit on dose rates is 100 miem per 7 days.

Analysis shows that the ISFSI design satisfies that requirement everywhere along the outer fence.

Calculated worst-case dose rates vary from 97 mrem /7 days (east and west outer fence centerlines) to 52 mrem /7 days (north and south outer fence centerlines) to 39 mrem /7 days (outer fence corners).

These dose rates include consideration for an improved HSM door' shield design which will reduce area dose rates below those predicted in the initial submittal of the Environmental Report.

When the proposed revision to 10CF'U 3 is-enacted, additional engineering and administrative controls will be evaluated and employed for the ISFSI in conjunction with other Calvert Cliffs site areas, as required to ensure compliance with the proposed revision to the 10CFR20 limit of 100 mrem /yr.

BGE001.0024.03

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i RBSPONSE TO NRC C010 TENTS.ON TER CALVERT-CLIFFS NUCLEAR POWER =PLRNT ISFSI ER QUESTION:

.ER-8 Para. 5. 2.1, p. 5. 2-1 ( Also SAR para 7.4.2, p. 7. 4-1 and Tigure 7.4-1).

(See also SAR Questions 7.0-11).

Provide a similar-figure for the dose rate vs. distance from the and of the module rows.

Also provide a relative dose rate decay curve.

RESPONSE:.

The response to question ER-3 contains the plots of dose rate" vs., distance from the HSMs.

l The relative dose rates at any casired time can be readily i

' derived-from theLattached gamma source term and neutron source ~

term decay l curves.

The-dose rates are proportional 1.to the neutron.and gamma source terms'and will decline similarily.

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RESPONSS TO 3fRC COBOtENTS ON TER CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI ER (Continued)

QUESTION:

ER-8 (Continued)

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4' RESPONSE TO NRC COIGtENTS ON TEE-CALVERT CLIFFS NUCLEAR POWER 71JGIT ISFSI ER (Continued)

QUESTION:

ER-8 (Continued)

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I RESPONSE TO NRC COMMENTS ON TNE CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI ER 1.

l' QUESTION:

ER-9 Para. 5.2.1 p.

5.2-1 (clso SAR Section 7.6,

p. 7.6-1).

SAR Section 7.6 does not provide an estimated off-site collective dose assessment due to direct and scattered radiation from the ISFSI.

Provide an offsite collective dose assessment for the population within two miles of the site and compare the results to the collective dose to the same population from.

reactor operations.

RESPONSE

Table 2.1-1 of the Environmental Report lists the population distribution within 10 miles of the Calvert Cliffs site.

1 Within the 0-2 mile radius, the population is 215 persons, l

z with no residences within the 0-1 mile radius.

Using the calculated dose rate from both direct and scattered radiation of.1.5E-5mR/h at 1 mile, the collective dose from the fully L

loaded ISFSI is 28.25 person-millirem.

This is less than 1%

L of the collective dose from reactor operations (13.5 mrem /yr x l

215) of 2902.5 person-millirem (SAR Section 7.6.2 Ref. 7.10).

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BGE001.0024.03

RESPONSE TO NRC COMMENTS ON THE CALVERT CLIFFS NUCLEAR POWER PLANT 18FSI ER QUESTION:

ER-10 Para.

5.6.4, p. 5.6-1 (also SAR Para.

7.2.2, p. 7. 2-1). This section of ER is in apparent conflict with the SAR.

That is, "the only postulated mechanism, for release of the radioactive material is the... non-fixed surf ace contamination on the DSC exterior."

Yet the ER states, "Any condensation-

. that may form on the DSC surface will not be contaminated since the DSC will not be contaminated."-

What is the potential for and possible pathway for the transport of contaminated rainfall and condensate runoff from the DSC?

What holdup and monitoring, if any is planned?

i

RESPONSE

The statement in the ER Para.

5.6.4, p. 5.6-1 is based on the technical specification on the DSC non-fixed surface contamination (SAR Para 10.3.2.6, p.10.3-9). This technical e

specification states that "smearable surface contamination 2

levels shall' be less than 2200 dpm/100 cm from beta and gamma L

2 emitting surfaces and 220 dpm/100 cm from alpha emitting surfaces". These non-fixed surface contamiriation levels are consistent with' the requirements of 10 CFR 71.87 (1) (1) and 49

. CFR 17 3. 4 4 3 which. regulate ' the use of spent fuel-shipping

-containers traveling on public rights-of-way. Consequently,

-these contamination levels are considered acceptable for exposure to the environment without further requirements for monitoring.

Hence the DSCs with _ non-fixed contamination levels below those given-above are described-as "not-contaminated" in the ER.

The ER discussion in para.

5'.6.4

.will be clarified by deleting ~the term "not contaminated" to

-describe the DSC surface condition and adding a description consistent with SAR Para.10.3.2.6 regarding the potential DSC '

i surface contamination levels and ' their acceptability for exposure to the environment.

Further, the potential for rainfall or atmospheric condensation dissolving or entraining contaminants on the 1

s surface of the DSC is negligible.

Any storm-driven rainfall:

which penetrates the weather protection.on the HSM air. vents will not fall directly.onto the DSCs but wi'll' trickle down the-HSM tinside, walls and bypass the DSC.

The DSC surface temperatures resulting from the decay heat of the spent fuel will prevent formation of any condensate on DSC surfaces.

However, even if.there is condensation, the DSC contamination

. levels are considered acceptable,for ' exposure to the environment per requirements of' 10 CFR71.87 (i) (1) and 49 CFR 173.443, no holdup and monitoring of runoff from the DSC is necessary.

BGE001.0024.03

RESPONSE TO NRC COMMENTS ON THE CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI ER QUESTION:

ER-11 Para 6.1.2 p.

6.1-3 (SAR para 7.3.4 p.

7.3-1).

What additional dosimetry will be added to monitor the area surrounding the ISFSI?

Describe the type,

number, and location of monitors that will be used to record dose rates along the ISFSI fence and instrumentation used for the CCNPP environmental monitoring program to be placed around the ISFSI.

Also, please describe the health physics procedures l

and instrumentation to be used to monitor dose rates in work l

areas around the transfer cask and HSMs during handling and I

transfer o,, trations to assure that occupational exposures are l

ALARA.

RESPONSE

The.SFSI Safety Analysis Report estimates the environmental l

radiological impact due to routine operation of the Spent Fuel Storage Facility and due to the design basis accidents at the facility. The expected environmental ef fect i

i1 and it is well monitored by the CCNPP operational mon!.t..uig nettlork presently in service.

-The present radiological monitoring program wil.1 be complemented as follows:

a.

Sixteen thermoluminescent dosimeters (TLDs) will be placed in 16 geographical sectors around the facility.

TLDs will be read monthly.

l b.

Two pressurized ionization chambers (PICS) will-be placed at the northeast (most prevalent wind

'l I

direction) and southwest corners of the storage facility.

These detectors are provided with continuous graphical recording which can be surveyed daily directly and/or by using a

telemetric system.

j c.

Two air samplers will be collected with the two PICS at the ISFSI and will be sampled weekly.

d.

Two monitoring sites, presently in operation and located in the vicinity of the facility, will be upgraded-by adding one air sampler and a TLD (at the principal meteorolcgical tower), and one PIC and a TLD (at the Camp Conoy vegetable garden).

BGE001.0024.03

-1 l

... ~.. _...

. 6 i.

RESPON8E TO NRC CONNENTS ON THE i

CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI ER (Continued)

QUESTION:

ER-11 (Continued)

Vegetation and soil samples near the facility will e.

be collected quarterly.

The present radiologichl environmental monitoring program, and additional 8 TLDs, one air sampler, and one PIC will assure the preoperational monitoring around the ISFSI.

Calvert Cliffs health physics procedures will be used during the loading and transfer operations.

An~ALARA review of-each operational milestone will be performed to assure that occupational exposure is kept ALARA.

t.

i i

BGE001.0024.03-w.-,

/(

1 f'

w.

s RESPONSE TO NRC COMMENTS ON TEE CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI ER QUESTION:

ER-12 a.)

Para 7.1.1, p. 7.1-1.

(See also SAR para 8.2.8, p 8.2-10

& -11).

To what degree are particulates contained?

b.)

Provide the isotopic inventory source term for the design 1

basis fuel assembly used in this accident analysis.

RESPONSE

a.)

The fo*Alouing release mechanisms were evaluated in the e

manner of SAND 80-2124 :

impact rupture,-burst rupture, 1

i diffusion, leaching, oxidation, and crud-release. It was concluded that noble gases may escape by diffusion and no other release mechanisms were probable.

1 Two systems for radionuclide confinement are novided by the NUNOMS*-24P system:

fuel rod claddk.y confines the fuel and fission products; and the DSC. 'contains the.

contents of the fuel assemblies and.the crud' adherent to the fuel. rods. Under normal conditions, both confinement -

1 features are. intact.

If any of the fuel rod. cladding.

f ails during storage,.the 'DSC.will become pressurized with a-mixture of. helium (the DSC fill gas) and fission 1

gases.

It is noted that the criteria for fue) cladding

~

' temperature in storage is established based on a

probability of failure of the peak temperature fuel. rods of 'less than 0.5%

(PNL-6189)2 In the. postulated accident, the DSC-is assumed to be _ pressurized by f allure of 100% of all: fuel rods, worst-case fission gas release fractions,- and: anc elevated temperature equal-to the worst-case thermal accident conditions.

The ' DSC is

' designed.and-shown by structural ~ analysis to withstand this pressure ' with a substantial margin ~ of. safety.

Additionally, the'DSC closure welds.are fully redundant, j

and' both are welded and inspected'to the' standards of the :

ASME Code.

While the release of 100% of the cladding gap; fission. gas-is assumed E for the purpose.of accident assessment, no mechanistic release path' exists.,

1Wilmot, E.L.,-. " Transportation-Accident Scenarios for CommercialDSpent. Fuel,"

Transportation Technology' Center, Sandia j

National Laboratories, February 1981, SAND 80-2124.

2Leyy, 7, 3,,.et al, " Recommended Temperature Limits for Dry Storage of1 Spent Light Water Reactor Zircaloy-Clad Fuel Rods in L

Inert-Gas," PNL-6189, May 1987.

BGE001.0024.03 g

l

n (16 l

RESPONSE TO-NRC ColOtRNTS ON THE D

CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI ER I

(Continued) l-QUESTION:

ER-12 (Continued)

Diffusion of the noble fission gas through the canister shell is assumed to occur.

The release of particulate material would require gross breach of the DSC pressure boundary.

Because of the ductile nature of the canister materials, the quality of construction inherent in the ASME Code, and the conservative nature of the design described above, a breach of sufficient size to cause release of radioactive particulate matter is incredible.

l Therefore, complete particulate confinement is assumed in l

the referenced accident analyses.

.i 3

=b.)

The OCRWM Database.was used as a basis for the' gross fuel radionuclide source-term.

A lengthy-list of the-l nuclides present in one metric ton of heavy metal was i

extracted from the database and has been attached to this L

response.

The data were 'obtained using ORIGEN calculations for. 8 year cooled, 45,000 mwd /MTU burnup PWR fuel.

The values in the " Curies" column should be

. multiplied by (0.386 MTHM

24 assemblies /DSC) to obtain the total source term per DSC.

s

.i 3" Characteristics of Spent Fuel, High Level Waste, and other Radioactive Wastes Which May Require Long Term Isolation", Office of Civilian Radioactive Waste Management,' DOE /RW-0184, December, 1987.

BGE001.0024.03 I

f

1_

I i

RESPONSE TO NRC COMMENTS ON THE CALVERT CLIFFS NUCLEAR POWER PLANT 18FSI ER (Continued)

QUESTION:

ER-12 (Continued) i

Argon, krypton, and radon are the only noble gases available for release.

Kr-8S-was determined to be the only significant contributor, as confirmed by Elias and Johnson.

A release fraction of 2.1% was calculated i

using the methodology of ANSI /ANS-5.4-1982.

100% of l

5 the " gap activity" was presumed to be available for release into the DSC cavity.

Furthermore, the-DSC release fraction was assumed to be 100%.

The amount of Kr-85 used as a radiological source term for one ruptured DSC was therefore:

t Cl MTHM A887 0 - 7.13E+ 03 X0.386 X 24 X 0. 021 MTHM Assy DSC O

- 1. 3 9E+03 g

t l

l

'Elias, E.,

Johnson, C.B.,

" Radiological Impact of Clad and.

4 Containment Failures in At-Reactor Spent Fuel Storage Facilities,"

Electric Power Research Institute, EPRI NP-2716, October 1982.

5ANSI /ANS-5.4-1982, " Method for Calculating the Fractional Release'of Volatile Fission Products from Oxide Fuel."

BGE001.0024.03

.1 l

i RE8PONSE TO NRC COMMENTS ON THE i

CALVERT "di4FF8 NUCLEAR POWER PLANT ISF81 ER j

(Continued)

QUESTION:

ER-12 (Continued)

Sperit Fuel Repository Characteristics Date Base Developed by: Dek Ridge Nationel Laboratory, Oak Rldge, TN.

Type of Reactor: PWR 45,000 Elepsed Decoys 8 years All isotopes rep *esenting: All nuclides Percentage i

Isotope Curles of Total N 3 6.25E+02 0.109 %

BE 10 6.66E 06 0.000 X C 14 1.69E+00 0.000 %

b SI 32 2.57E 07 0.000 %

P 32 2.5?E 07 0.000 X

,i s 35 2.40E 09 0.000 1 CL 36 1.30E 02 0.000 %

AR 39 8.53E 05 0.000 %

AR 42 4.64E 13 0.000 %

K 40 5.90E*09 0.000 %

K 42 4.64E 13 0.000 %

CA 41 2.08E 04 0.000 %

CA 45 1.53E 06 0.000 %

'3 SC 46 1.11E 10 0.000 %

v $0 1.11E 14 0.000 %

MN 54 1.24E+00 0.000 %

FE 55 6.18E+02 0.108 %

FE 59 9.34E 18 0.000 %

C0 58 2.55E 09 0.000 %

C0 60 2.87E+03 0.502 %

NI 59 4.76E*00 0.001 %

N! 63 7.08E+02 0.124 %

i ZN 65 2.59E 02 0.000 %

~

SE 79 5.53E 01 0.000 %

KR 81

'7.17E 07 0.000 %

KR 85 7.13E+03 1.247 %

)

RB 87 ~

2.89E 05 0.000 %

I

$R 89 2.78E 12 0.000 %

SR 90-7.97E+04 13.934 %

I Y 90' 7.97E+04 13.934 7.

l Y 91 8.66E 10 0.000 %

j 2R 93 2.61E+00 0.000 %

2R 95-2.49E 08 0.000 %-

NB 93M 1.02E+00 0.000 %

N8 94 1.62E+00 0.000 %

NB 95 5.72E 08 0.000 %

1 N8 95M 1.85E 10 0.000 %

l

't Mo 93 3.20E 02 0.000 %

1 TC 98 9.47E 06 0.000 %

TC 99 1.70E+01 0.003 %

RU103 6.00E 17-0.000 %

RU106 2.26E+03-0.395 %

RM102 2.75E 01 0.000 %.

RM106 2.26E+03 0.395 %

P0107.

1.54E 01 0.000 %

AG108 1.27E 03 0.000 %

AG108M 1.43E 02 0.000 %

A0109M 1.57E 02 0.000 %

AG110 2.12E 02 0.000 X AG110M 1.60E+00 0.000 %

CD109 1.57E 02 0.000 %

C0113M 5.33E+01 0.009 %

BGE001.0024.03

~

I RESPONSE TO WRC COMMENTS ON THE CALVERT CLIFFS WUCLEAR POWER PLRNT ISFSI ER (Continued)

QUESTION:

ER-12 (Continued)

Spent Fuel Repository Characteristics Data Base Developed byl Det klope National Laboratory, ook Ridge, TN.

(Contirwed)

Percentope isotope Cuties of total C0115M 3.461 17 0.000 %

IN113M 1.56t 05 0.000 %

IN114M 2.72E 16 0.000 %

IN115 1.57t 11 0.000 %

$N113 1.5H 05 0.000 %

$N119M 1.38t+00 0.000 %

$N121M 8.87t*01 0.000 %

$N123 5.12t 04 0.000 1

$N126 1.04t+00 0.000 %

$8124 4.37t 12 0.000 %

$8125 2.40t+03 0.420 %

st126 1.45t 01 0.000 %

$4126M 1.04E+00 0.000 %

ft123 5.23t 12 0.000 %

18123M 1.26t 06 0.000 %

it125M 5.86t+02 0.102 %

ti127 1.05t 04 0.000 %

it127M 1.071 04 0.000 %

1 1129 4.23t 02 0.000 %

C$134 1.54t+04 2.692 %

Cs135 5.66t 01 0.000 %

C5137 1.15t+05 20.105 %

l BA137M 1.09E*05 19.056 %

LA138 2.24t 09 0.000 %

Ct142 3.67t 05 0.000 %

CE144 7.94t+02 0.139 %

j PR144 7.94t+02 0.139 %

PR144M 9.53t*00 0.002 %

N0144 2.18E 09 0.000 %

PM146 1.52t+00 0.000 %

PM147 1.58t+04 2.762 %

PM148M 1.55t 17 0.000 %

sM146 5.16t 07 0.000 %

$M147 5.15t 06 0.000 %

'.M 143 8.02t 11 0.000 %

$M149 9.82t 13 0.000 %

SM151 4.39t + 02 0.077 %

EU150 2.95t 05 0.000 %

EU152 7.51E+00 0.001 %

EU154 8.45E+03 1.477 1 EU155 3.34t+03 0.564 1 C0152 6.06t 13 0.000 %

G0153 1.83t 02 0.000 %

18160 9.08E 10 0.000 %

H01664 4.87t 03 0.000 %

iM170 1.29E 08 0.000 %

TM171 8.78t 05 0.000 %

j 6

(U176 3.24t 11 0.000 %

LU177 2.83t*09 0.000 %

LU177M 1.23E 08 0.000 %

Hf175.

2.60E 12 0.000 %

MF181 9 76t 19 0.000 %

HF182 3.74E 07 0.000 %

TA182 1.43E 06 0.000 %

w)81 5.24t 08 0.000 %

BGE001.0024.03

4 RESPONSE TO WRC COMMENTS ON THE CALVERT CLIFF 8 NUCLEAR POWER PLANT IBF81 ER (Continued)

QUESTION ER-12 (Continued)

Spent Fuel Repository Cf4racteristics Date Base ces m.ed by Oak tio8e National Laboratory, Ook tid 8e, TN.

(Continued)

Percentate llotope Curles of fot4L TH2%

3.12t 01 0.000 %

PA231 3.27t 05 0.000 %

PA233 4.76t 01 0.000 %

PA234 4.06t 04 0.000 %

PA234M 3.12t 01 0.000 %

U232 5.79t 02 0.000 %

U233 3.68t 05 0.000 %

U234 1.34t+00 0.000 %

U235 1.85t 02 0.000 %

U236 3.52E 01 0.000 %

U237 2.79t +00 0.000 %

U238 3.12t 01 0.000 %

U240 8.44t 07 0.000 %

WP235 6.42E 05 0.000 %

WP236 1.02t 05 0.000 %

NP237 4.76t 01 0.000 %

hP238 7.43t 02 0.000 %

kP239 3.08t+01 0.005 %

WP240M 8.44t 07 0.000 %

PU236 1.60E 01 0.000 %

PU238 4.40t+03 0.769 %

PU239 3.59t+02 0.063 %

Pv240 5.80t+02 0.101 %

PU241 1.14t+05 19.930 %

PU242 2.64t + 00 0.000 %

'U243 3.51t 07 0.000 %

PU244 8.45E 07 0.000 X P't246 2.67t 14 0.000 %

AM241 2.01t+03 0.351 %

AM242 1.48t+01 0.003 %

AM242M 1.49t+01 0.003 %

AM243 3.08t+01 0.005 %

AM244 8.46t 11 0.000 %

AM245 1.04t 10 0.000 %

AM246 2.67E 14 0.000 %

CM242 1.26t+01 0.002 %

CM243 3.38t+01 0.006 %

CM244 3.30t+03 0.577 %

CM245 4.23t 01 0.000 %

CM246 1.02t 01 0.000 %

CM247 3.51t 07 0.000 X CM248 9.75t 07 0.000 %

CM250 1.07t 13 0.000 %

8K249 7.21t 06 0.000 %

8K250 2.84t 11 0.000 %

CF249 1.16t 05 0.000 %

CF250 3.81t 05 0.000 %

CF251 3.50E 07 0.000 %

CF252 8.91t 06 0.000 %

E ES254 2.83

...................................E 11 0.000 %

Subtotal Curies.

5.73t+05 100.000 %

Total aLL isotopes.

5.72t+05 BGE001+0024.03 l

i RESPONSE To NRC COMMENTS ON TEE CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI SAR l

Section 7 QUESTION:

7.0-1 Paragraph 7.1.1, page 7.1-1.

With respect to the Performance Improvement Plan, what is the status of the Calvert Cliffs Radiation Safety Manual, ALARA Program, and Radiation Safety (Implementation) Procedures in relation to the ISFSI project?

RESPONSE

With respect to Radiation

Safety, the Calvert Cliffs Performance Improvement Plan (PIP) addresses generic site management and process control improvements programs. The PIP does not specifically address the Calvert Cliffs Radiation Safety

Manual, ALARA Program and Radiation Safety i

(Implementation)

Procedures (RSPs) except as part of the Procedures Upgrade Program (PUP).

As of August 1, 1990, 43%

of the RSPs have been upgraded to current administrative standards with the remainder scheduled to be completed by the end of 1991.

Since the PIP, as a whole, is directed at improving management assessment process control and verification, the Radiation Safety Program will be strengthened on an overall basis.

Present RSPs adequately address radiological aspects of the ISFSI project construction and operation.

I l

1 i

l BGE001.0024.03 1

RESPONSE TO NRC COMMENTS ON THE CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI SAR Section 7 QUESTION:

7.0-2 Para 7.1.2, p.

7.1-2.

Provide an evaluation of the possible long term storage effects of using borated water instead of demineralized watering filling the DSC cavity.

Specifically, address the potential corrosion effects.

RESPON9Et The cavity of the DSC will be submerged in the spent fuel pool for about 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and, on removal from the pool, will contain borated water from the spent fuel pool for less than 50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months .

There is a substantial body of industry experience with exposure of austenitic stainless steels to borated water since that condition exists in most PWR spent fuel pools.

A literature search did not reveal any journal articles referring to corrosion of austenitic stainless in pools, from which one could infer that none has been observed since anecdotal experiences have all been very good.

A combustion Engineering study on the effects of borated water on corrosion of low alloy eteels reports a complete absence of corrosion in a very aggressive environment (dripping borated water and wet borated steam) for Type 304 and other corrosion resistant materials.

The author concludes that "... corrosion resistant alloys such as Types 304 are not susceptible to borated water corrosion.

Furthermore, stressed samples of these materials did not exhibit any localized forms of corrosion, such as stress corrosion

cracking, hydrogen embrittlement, etc." (Reference 1) l After the DSC cavity has been drained, about 2 to 4 gallons of residual borated water will remain due to surface tension. As the borated water evaporates during the vacuum drying process, the ortho-boric acid crystals will precipitate out of the solution at concentrations substantially higher then 2000 ppm.

After the free water has ' evaporated, a small amount of dehydrated B0 crystals will remain in the cavity.

A 23 literature search did not uncover any reference to corrosive or aggressive behavior of anhydrous boric acid and discussions with chemists at an industry supplier (U.S. Borax) revealed that it will only become corrosive in the molten state.

The melting temperature for anhydrous boric acid is about 450*C, well above the peak DSC material temperatures and the peak BGE001.0024.03

RESPONSE TO NRC C0KMENT8 ON THE CALVERT CLIFF 8 NUCLEAR POWER PLANT ISFSI SAR (Continued)

Section 7 QUESTION:

7.0-2 (Continued) fuel cladding temperaturca.

Even if corrosion of the DSC shell or basket is postulated, the extremely small quantity of borate and the expected corrosion mechanism of general surface oxidation is not likely to lead to degradation of the structural integrity of the DSC shell or basket.

Reference:

1.

J.F.

Hall, " Corrosion r.? Low Alloy Steel Fastener Materials Exposed to Borated k cer",

Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, Ed. by G.J.

Thoue and J.R. Weeks, The Metallurgical Society, 1988, i

BGE001.0024.03

l RESPONSE TO NRC CONNENTS ON THE CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI SAR Section 7 QUESTION:

7.0-3 Para 7.1.3, p.

7.1-3.

The second and third paragraphs about maintenance appear to be contradictory.

Please clarify.

What maintenance activities are foreseen?

RESPONSE

The third paragraph is a restatement of BG&E's' commitment to use proper ALARA procedures for any ISFSI unplanned maintenance activities performed in a radiation field, should it ever become necessary.

No such activities are foreseen.

The second paragraph points out that planned maintenance activities on the operational equipment will be performed in unrestricted areas. Since the third paragraph has resulted in confusion, it will be deleted.

Planned maintenance activities are preventive in nature and include motor oil changes, hydraulic oil filter replacement and the like.

i l

l l

l '

l l

l BGE001.0024.03 l

l

RESPONSE TO NRC CONNENTS ON THE CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI SAR Section 7 l

QUESTION:

7.0-4 Paragraph 7.4.2, p.

7.4-1.

How does BG&E plan to restrict access to the ISFSI area and transport pathway for the DSC to the ISFSI since the ISFSI site is not within the plant protected area and there is access from visitors to Camp Conoy?

Of greatest interest is the time period when the l

seismic restraint is being installed and the dose rate at 1 foot from the DSC is 1790 mr/hr.?

What is the maximum dose rate during this time period at the ISFSI fence?

Demonstrate comoliance with 10CFR20.105.

RESPONSE

If the initial DSC transfers are performed during camp Conoy hours, visitor access will be restricted and controlled by i

BG&E Security.

During the DSC seismic restraint installation, radiation dose rate measurements will be made by Radiation l

Safety Technicians at selected locations, to supplerant environmental surveillance stations and thermoluminescent dosimeters and to validate calculated dose rates.

During the seismic restraint installation, the dose rate at the ISFSI fence is determined to be 1.96 mrem /h in the calculation below.

Since the seismic restraint installation is conservatively estimated to require 0.08 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to complete, the resultant dose at the fence (1.96 mrem /h*0.08h) is 0.16 L

mrem without credit for the DSC transfer cask which in fact shields the majority of the source area.

Given that the maximum dose rate at any location on the ISFSI fence from all l

five phases of the ISFSI is 0.58 mR/h, the maximum total dose rate at the fence due to seismic restraint installation is l

estimated as 0.74 mrem (0.16 mrem + 0.58 mrem) in one hour.

This is well within the 10CFR20.105 requirement that no individual (member of the public) may receive greater than 2 mrem in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

The time duration of 0.08 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (5 minutes) to install the seismic restraint is considered conservative.

Experience at 8 ISFSI has demonstrated the Duke Power Company's Oconee NUHOMS installation time to be less than 0.017 hour1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months (1 minute).

1.

BGE001.0024.03

s i

RESPONSE TO NRC CONNENTS ON TER CALVERT CLIFF 8 NUCLEAR POWER PLANT ISFSI SAR (Continued)

Section 7 QUESTION:

7.0-4 (Continued)

DOSE RATE AT rnC !*FSI SECURITY FENCE RESULTING FROM THE DSC 1h W OPEN HSM DOOR The maximum contact dose rate on ti.* DSC bottom and surf ace is 1955 mram/h!.

The minimum distance between the DSC surface and an individual at the closest approach to the outer ISFSI security fence (east / west portion of the fence) is estimated to be 66.7 feet.

To calculate the spatial dose attenuation at this distance without taking credit for attenuation or buildup in air the following equation is used.

Assume the DSC end surface is a disc source of radjas R,

emitting S particles /sec isotropically. The flux 6(r) at a point P with distance X from the disc source is given by (Reference 1):

$ ( P) -

In (1 +

)

Assuming that the flux distribution is semi-isotropic (outwardly directed) on the DSC end surface, the flux at point P will be

$ ( P) -

In(1+

)

f BGE001.0024.03 l

l l

RESPONSE TO NRC COMMENTS ON TEE CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI SRR (Continued)

Section 7 QUESTION:

7.0-4 (Continued)

The outside diameter of the DSC is 66.25" (Radius = 33.625").

Substituting the values of X and R in the above equation, the flux at point P will be 6(P) = 0.001 S.

The flux is directly proportional to the dose rate, so the dose rate at point P will be 0.001 times the dose rate at the DSC surf ace. The dose rate at point P on the fence due to the visible DSC and surface will be 1955*0.001 = 1.96 mram/hr.

Note that at 66.7 feet from the DSC surface an individual at the outer security fence can only see a fraction of the DSC end surface since the DSC is partially shielded by the cask and HSM front wall.

For these calculations it is conservatively assumed that the entire DSC bottom end surface is visible.

Reference:

1.

Lamarsh, John R., " Introduction to Nuclear Engineering",

Addison Wesley Publishing Co.,

1977.

BGE001.0024.03 i

mare.-

__i_.-o-r#-

,a

ww w--

RESPONSE TO NRC C0bOLENTS ON THE CALVERT CLIFPS NUCLEAR POWER PLANT ISFSI SAR Section 7

-QUESTION:

7.0-5 3

Para 7.4.2, p. 7.4-1 This section indicates that daily visual inspection of the CCNPP ISFSI HSM air inlets and outlets can be performed by keeping inspection personnel at least 32 feet away from the HSM face.

This may not be possible for effective visual inspection of the " hidden" inner rows of HSMs.

Perform a revised calculation of the inspection personnel done with a more realistic time-distance scenario for complete visual inspectio~n of the HSM inlets and outlets in the ISFSI, and suggest a plausible method for verifying that the air inlets and outlets of inside rows are free of foreign materials.

RESPONSE

The subject SAR Paragraph 7.4.2 will be revised to state:

"The annual HSM air inlet vent inspection dose is estimated to be 46 mram/yr.

This value is derived by assuming that one inspector performs a daily inspection walking at an average speed of 3 mph on a path passing directly in front of all 120 air inlet vents of 'the ISFSI.

The distance between the inspector and each HSM front wall is assumed to be 36' (the

- half-distance between each row of modules).

It is further assured that all five phases (120 HSMs) are filled with design basis fuel.

No credit is taken for radioactive decay of the fuel during storage.

Remote cameras will be used for air outlet surveillance, therefore no exposure will result from their inspection."

Three items should be noted regarding HSM inspection and the resulting exposure estimates l

The 32' distance stated in the original submittal of the i

SAR corresponds to the half-distance between the HSMs in a preliminary version of the ISFSI layout.

The correct half-distance is 36'.-

This value is used as an assumption-for calculations and no implication is made that it is a requirement for HSM surveillance.

The 4

inspector-HSM distance is to be decided by'ths inspector based on good ALARA judgoment and the ability to detect blockage of the HSM air inlet vents.

BGE001.0024.03

t RESPONSE TO NRC CONNENTS ON TER CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI SAR (Continued)

Section 7 l

QUESTION:

7.0-5 (Continued) e The air inlet vents are one foot high by six feet wide.

The inspector will have an unobstructed view of each HSM as he passes. Visual inspection from any distance within the " alley" between HSMs is sufficient to reveal blockage 1

of the vents or damage to the air inlet screens.

The inspection dose estimate has been revised from 88 to 46 mrem /yr.

This results from an improved HSM door design which reduces HSM door surface dose rates from 66.0+11.0 mrum/hr (neutron + gamma, respectively) to 21.7+2.1 mrem /hr.

Other HSM surface dose rates are as reported in the SAR.

Both values were calculated by modeling the HSM surfaces as isotropically emitting surface sources and integrating the dose received as a function of time along the path of the inspector shown below. The contribution from skyshine, or air-scattered, radiation was included as well.

72' -

- 36' c

(- -

7 7

r- - 1

~ --

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(

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4 1

4 l

l L _.J L _

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1Even at a distance of 72', a 1" object will subtend a viewing angle of 3.9 min, which is more'than sufficient for a near-100%

probability of deteccion.

(Van Cott, Kinkade, " Human Engineering Guide to Equipment Design," American Institutes for Research, l'

1972.)

BGE001.0024.03

RESPONSE TO NRC ColOLENTS ON TER CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI SAR Section 7 QUESTION 7.0-6 Tables 7.2-1, 7.2-2.

The assumed source term of 24 CCNPP fuel assemblies in a DSC has an 8% smaller total gamma component, but a 44% higher neutron component than the NUTECH TR.

Provide specific design details of the TC, DSC, and HSM which have been modified to account for this greater source and provide sufficient additional shielding.

RESPONSE

Several changes to the generic NUHOMS'-2 4 P design neutron shielding were required in order to accommodate the higher neutron source term.

The changes to neutron shield thicknesses are summarized below.

Other changes to the shielding design thicknesses' vere also made as indicated in the CCNPP ISFSI SAR.

Neutron Shield UHOMS'-2 4 P CCNPP NUHOMS' Go aric Design ISTSI component Direction Desion Transfer Cask Top Axial 2.00" NS-3 3.00" NS-3 Transfer Cask Bottom Axial 2.50" NS-3 3.50" NS-3 Transfer Cask Radial 3.00" Water 4.00" NS-3 HSM Door (Axial) 2.00" NS-3 10.75" concrete' HSH Radial 36" Concrete 36" concrete DSC All None None

'This HSM door. design is an improvement of the one originally presented in the initial submittal of the CCNPP ISFSI SAR.

BG&E will revise the SAR to incorporate this design.

l BGE001.0024.03 l

[

RESPONSE TO NRC C030 TENTS ON TER CALVERT CLIFFS NUCLEAR POWER PLhWT ISFSI SAR Section 7 QUESTION:

7.0-7 Table 7.2-2.

Explain why the source gamma energy spectrum in Table 7.2-2 has no gamma source for energy groups 4, 9,

and 12.

RESPONSE

The zero source term for gamma groups 4, 9, and 12 is a result of the methodology used to convert the gamma source term from the ORIGEN2. energy structure to the CASK energy structure.

ORIGEN2 'was used to develop the source terms.

It is restricted to its own internal 18 gamma ray energy' group structure.

Shielding calculations were performed using ANISN and the CASK 189-22n cross section set.

The ORIGEN2 sources must be converted to the CASK energy structure in order to' perform shielding calculations. Although both structures have 18 groups, there is not a 1-1 correspondence in energies.

As a result of overlapping between the ORIGEN2 and CASK groups, some CASK groups were left with a zero source term.

Note that gsmma power (MeV/sec), and not y/sec, was used as a basis for the conversion; therefore the effect of the zero source groups i

is negligible.

l l

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BGE001.0024.03 L

)

RESPONSE TO WRC COMMENTS ON TER CALVERT CLIFFS WUCLEAR POWER PLANT ISFSI SAR Section 7 QUESTION:

7.0-8 Table 7.3-1.

Two dose rates in Table 7.3-1 of the CCNPP SAR are significantly greater than in the NUTECH TR.

The doses are 1808 mr/hr in the center of doorway of the DSC in the HSM and 2527 mr/hr in the dry gap edge of the DSC cover plate.

Explain why these dose rates are much higher than in the NUTECH TR and what shielding alternatives were investigated to reduce these dose rates.

Provide calculations which support the dose rates presented in Section 7.

RESPONSE

The CCNPP HSM doorway dose rate (1808 mram/hr) is higher than the corresponding generic NUHOMS -24P dose rate (760 mrem /hr) due to des: gn differences in the fuel to be stored and the DSC shield geometries.

Both values were calculated using the methodology presented in the NUHOMS*-24P Topical Report.

A number of different combinations of DSC, transfer cask, and HSM door shield thickness combinations were evaluated and the present design was determined to be acceptable in terms of the operational, regulatory, and ALARA objectives.

Note that the only operation which takes place in the HSM doorway radiation field is the seismic restraint installation.

The CCNPP NUHOMS' seismic restraint design includes refinements which allow the restraint to be more easily handled and quickly

,placed:than the generic design.

.Since-the remainder of the operational-and storage term radiological exposure rates are heavily intluenced by_.the HSM door exterior dose rates, an improved HSM door design has been developed for the CCNPP ISFSI which reduces the HSM door centerline dose rate from 77-to.24 mrem /hr.

The resulting dose rate is lower than that calculated for the TR.

The improved HSM door design will.be included in a revision of the CCNPP ISFSI SAR.

The dose rate in the dry annulus between the DSC cover plate and the cask for the generic NUHOMS'-2 4 P design is not reported in the NUHOMS'-24P Topical Report.

The calculated dose rate is 3111 mram/hr for the generic NUHOMSo-24P design.

This is higher than the corresponding CCNPP NUHOMS' design due to differences in the fuel (self shielding and source terms) and the DSC shield plug geometries.

Both values-were calculated using the methodology presented in the NUHOMS'-24P Topical Report.

BGE001.0024.03

-~.

i s

RESPONSE TO WRC CONNENTS ON THE CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI SAR (Continued)

Section 7 QUESTION:

7.0-8 (Continued)

Calculations supporting Section 7 of the Calvert Cliffs ISFSI SAR will be provided I

BGE001.0024.03 l

RESPONSE To NRC ColOLENTS ON TER CALVERT CLIFFS NOCLEAR POWER PLANT ISFSI SAR Section 7 1

QUESTION:

7.0-9 Table 7.4-1.

The estimated occupational exposure for one HSM load has numerous differences when compared to its counterpart in the NUTECH TR.

These discrepancies include missing steps, fewer number of personnel for an operation, less time for an operation, and a greater distance from the cask during the operation.

Explain these apparently non-conservative discrepancies.

Also, explain why the two highest calculated dose rates from Table 7.3-1 do not appear in this assessment at all.

RESPONSE

The estimated occupational exposures'for one HSM load listed in-the SAR Table 7.4-1 are based on the experience gained during the fuel loads at the Carolina Power and Light Company's H.B.

Robinson NUHOMS'-07P ISFSI.

The estimated occupational exposures are also consistent with subsequent experience gained durin fuel loads at the Duke Power Ccmpany's oconee NUHOMS*g the

-24P ISFSI. The apparent differences between the NUTECH TR and Calvert Cliffs ISFSI are due largely to the regrouping of the various tasks listed in the SAR Table 7.4-1.

j The two highest calculated dose rates from the SAR Table 7.3-1 are 1808 mram/hr.at the center of the HSM doorway and 2527 mram/hr at the outer edge of the dry gap between DSC and cask.

The dose rate for the installation of the seismic restraint in the-HSM is calculated-to be 1790 mrem /hr. It is used for 1

occupational exposure calculated in the SAR Table 7.4-1 L

instead of tho'1808 mram/hr calculated at the center of the HSM doorway.

l The dose rate at the outer edge of the dry gap between DSC and cask'is also included in the appropriate operations listed in the SAR Table 7.4-1. The ambient dose rates of 130.6. mrem /hr and ' 66.7 mrem /hr listed in the SAR Table 7.4-1 include contribution from the dry gap between DSC and cask. These ambient dose rates are the area averaged dose rates at the given distances.

l L

BGE001.0024.03

=-

RESPONSE TO NRC COMMENTS ON THE CALVERT CLIFFS NUCLEAR POWER PLANT ISFSI SAR Section 7 QUESTION:

7.0-10 Table 7.4-1.

It is not stated in the SAR where the hydraulic ram controls are, except that "all controls are mounted in one trailer mounted control panel" (para 4.7.3.7, p. 4.7-5).

.The safety related concern.s the potential exposure of the RAM operator to radiation from the DSC through the ran access opening in the transfer cask.

Please confirm the ambient dose during this operation is truly that adjacent the transfer cask (13.7 mrem /hr) as shown in the table, or correct the table for the higher dose.

RESPONSE

The hydraulic ram controls are mounted on a trailer which is located to the side of the hydraulic ram.

To have access to the controls, the ran operator is located approximately 20' away from the cask ram access opening and off to the side such that only a small portion of the opening is visible during the.

ram operation.

The use of the general area dose rate at 13.7 mrem /hr is justified based on the small dose rate of this distance from the ram access opening and the relatively small contribution of the ram access opening to the general area dose rate.

These two points are discussed in detail in the following paragraphs.

The 13.7 mram/hr general area dose rate shown in SAR Table 7.4-1 is based on an average distance of 8 feet from the cask surfaces.

The total dose rate on the cask surface is 83.6 mics/hr.

(Note that the current calculated dose on the cask surf ace 3.s 83.6 mrem /hr instead of the 85 mrem /hr shown in SAR Table

".3-1.

The 83.6 mrem /hr value will be reflected in a revision to SAR Table 7.3-1. )

The dose rate at the ram access cp2ning is calculated to be 1951 mrem /hr.

The dose rate at 20

.I feet directly in front of the ram access opening is calculated to be less than 1 mrem /hr.

This dose rate is calculated using the methodology described -in the response to SAR Question 7.0-4 and substituting the dimensions of the ram access opening otreaming path.

BGE001.0024.03

RE8PONSE TO NRC COMMENTS ON THE CALVERT CLIFF 8 NUCLEAR POWER PLANT ISFSI SAR (Continued)

Section 7 QUESTION:

7.0-10 (Continued)

The relative contributions from the cask surface and the ram opening are roughly proportional to the product of the doce rate and the projected area, given that the d.4 stance to the two is approximately the same.

The annular ram access streaming path has an outer diameter of 10" and the ram itself has a diameter of approximately 6 ".

The streaming path 2

therefore has a cross-sectional area of about 45 in.

The 2

cask's projected area is 181" x 89" 16,100 in.

We can

=

estimate the ratio of exposure from the two cources as:

4Sint X 1951 mrem /hr p,,

16,100ln2 X 83. 6 mrem /hr

- 0.06 The contribution from the cask surface is clearly dominant.

Also, note that the ram controls operator will be more than 8 feet away from the cask surface which will reduce the general area dose below 13.7 mrem /hr.

For this analysis, the 13.7 mrem /hr is used conservatively.

BGE001.0024.03

i I

RESPONSE TO NRC CONNENTS ON TER CALVERT CLIFFS NUCLEAR POWER PLANT 18 FBI SAR Section 7 l

QUESTION:

7.0-11 Figure 7.4-1.

Explain why the ISFSI dose rate vs. distance curve in Figure 7.4-1 of the SAR does not

.Sp off as quickly as the analogous curve in the NUTECH TR af cne about 2000 feet.

3

RESPONSE

The Calvert Cliffs ISFSI is planned to store fuel in up to 120 HSMs.

The storage is arranged in five construction phases; each phase consisting of two 2x6 arrays of HSMs.

The generic design depicted in Figure 7.4-1 of the NUHOMS*-24P

)

Topical Report cannot be compared directly to the Calvert cliffs. design since it only shows dose vs. distance from a single 2x10 array of HSMs.

In addition to having 120% of the direct dose emanating from the HSM front surface (as compared to the generic design), the Calvert Cliffs ISFSI has 120/20 as much skyshine from the roofs, plus additional skyshine from rows of HSMs hidden behind the front row.

It is this extra skyshine component which is primarily responsible for the differences in the two curves.

Note that BG&E plans to revise the Calvert Cliffs SAR to incorporate revised dose vs. distance curves.

These revised curves will incorporate the effects of an improved HSM door-design which will substantially lower area dose rates.

In addition, the new curves are based on calculations using ANISN to accurately predict the scattering and absorbtion of direct neutron radiation.

This is an improvement in the manual point-kernel methodology used in the previous SAR.

calculations.

The resulting dose rates are calculated to be-relatively lower at great distances due to consideration for neutron absorbtion.

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BGE001.0024.03

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