ML18018A501

Forwards Response to Radiological Assessment Branch Draft SER Open Item 169 Re Shielding Against Neutron & Gamma Rays Into Containment from Annulus

ML18018A501
Person / Time
Site:	Harris
Issue date:	04/19/1983
From:	Mcduffie M CAROLINA POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
LAP-83-107, NUDOCS 8304210437
Download: ML18018A501 (14)

Text

REGULATOf INFORMATION DISTRIBUTION TEM (RIDS)

ACCESSION NBR: 8304210437 DOC ~ DATE: 83/04/19 NOTARIZED:

NO FACIL:50-400 "Shearon Harris Nuclear Power Pl anti Uni,t ii Car ol ina 50-401 Shearon Har ris Nuclear Power Planti Unit 2i Carolina AUTH,NAtlE AUTHOR AFFILIATION tdCDUFFIE,M ~ A, Carolina -Power 8 Light Co ~

RECIP ~ NAME RECIPIENT AFFILIATION DENTONgH ~ Re Office of Nuclear Reactor Regulationi Director

SUBJECT:

Forwards response to Radiological Assessment Branch draft SER Open Item 169 re shielding against neutron 8

gamma rays into containment from annulus.

DISTRIBUTI N CODE:

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TITLE: Licensing Submittal:

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CIRQUE, Carolina Power & Light Company APR 19 1983 SERIAL:

LAP-83-107 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NOS.

1 AND 2 DOCKET NOS. 50-400 AND 50-401 DRAFT SAFETY EVALUATION REPORT RESPONSES RADIOLOGICAL ASSESSMENT BRANCH

Dear Mr. Denton:

Carolina Power 8 Light Company (CPltL) hereby transmits one original and forty copies of the response to the Shearon Harris Nuclear Power Plant Draft Safety Evaluation Report (DSER)

CP&L Open Item No.

169.

Carolina Power 5 Light Company will be providing responses to other Open Items in the DSER shortly.

Yours very truly, PS/cfr (6642PSA)

Attachment M. A. McDuffie Senior Vice President Engineering 5 Construction Cct Mr. Seymour Block (NRC-RAB)

Mr. N. Prasad Kadambi (NRC)

Mr. G. F. Maxwell (NRC-SHNPP)

Mr. J.

P. O'Reilly (NRC-RII)

Mr. Travis Payne (KUDZU)

Mr. Daniel F.

Read (CHANGE/ELP)

Chapel Hill Public Library Wake County Public Library Mr. Wells Eddleman Dr. Phyllis Lotchin Ms. Patricia T.

Newman Mr. John D. Runkle Dr. Richard D. Wilson Mr. G.

O. Bright (ASLB)

Dr. J. H.'Carpenter (ASLB)

Mr. J. L. Kelley (ASLB) pool 83042i0437 8304i9 PDR ADOCK 05000400 E

PDR 411 Fayetteville Street

~ P. O. Box 1551 o Raleigh, N. C. 27602

OPEN ITEM No.

169 (DSER Section 12.3.2)

Shielding against neutron and gamma rays streaming into containment from the annulus between the RPV and the biological shield has been inadequately analyzed.

RESPONSE

TO OPEN ITEM No.

169 This is a Radiological Assessment Branch Open Item in the Draft Safety Evaluation Report (DSER),

and was previously presented as FSAR Safety Review Question 471.2.

Attachment I provides the response to Question 471.2.

ATTACHMENT I FSAR Safety Review Question 471.2

FSAR SAFETY REVIEW QUESTION 471.2 (FSAR Sections 12.2.1, and 12.3.1)

In Section 12.2.1, neutron and gamma streaming from the annulus between the RPV and the biological shield should be analyzed with respect to dose rate levels in containment where occupancy may be required.

Section 12.3.1.11(c) states that a 3" neutron shield will be used to reduce this neutron streaming.

Please specify the neutron and gamma dose equivalent rates that will exist at specific locations within the various levels of containment prior to shield installation and after the shield is installed (i.e. what is the effective factor of reduction for gamma and neutrons of the installed shield).

A figure or table showing respective dose rates would be a suitable format.

Please use relevent experience, as necessary, to demonstrate that your proposed shielding would accomplish the objective of Regulatory Guide 8.8 Section C.2.b, namely that it will provide sufficient shielding to achieve ALARA exposure to occupants in containment while the reactor is at power.

Please specify the frequency at which entries are made into containment, the number of people making these entries and their stay time.

RESPONSE

TO QUESTION 471.2 The potential for radiation streaming (neutron and gamma) through the annulus around the reactor pressure vessel has been analyzed to determine the prevalent radiation fields during normal operation in areas of containment which may require occupancy.

Since operating experience (Reference 12.3.2-D, 12.3.2-E and 12.3.2-F) indicated that streaming gamma dose rates during normal operation are a

relatively small fraction of the corresponding neutron dose rates, the analysis of the streaming dose rates in containment has been limited to the determination of neutron dose rates.

The angular neutron flux as a function of energy which emerges from the surface of the reactor vessel has been calculated using the DOT-III (Reference 12.3.2-B) discrete ordinates code utilizing an S8 angular quadrature and a P3 Legendre expansion for anisotropic scattering.

The vessel emergent angular fluxes have been used as input to a Monte Carlo analysis of the streaming neutrons.

Morse-CG (Reference 12.3.2-4),

a general purpose Monte Carlo multigroup neutron and gamma ray transport code with combinatorial geometry, has been used to compute the neutron streaming and the resultant dose rates on operating deck and various other locations inside containment.

The cross section library used in the calculations is based on DLC-23 or CASK library (Reference 12.3.2-A).

This library is a coupled neutron and gamma ray library and the data in the library are obtained by collapsing cross sections over a PWR core spectrum.

The three dimensional geometrical model used for the dose rate calculations includes the reactor vessel steel shell, containment structural concrete, major features of the containment internal structures such as the refueling cavity, the shield walls around steam generators and

pumps, a detailed

(Refer ence 12.3.2-F) scaled to full power indicate that neutron dose rates at locations where predicted SHNPP dose rates are approximately 8.7 REM/HR, range from 60 to 65 REM/HR.

Dose rates on intermediate elevations in these plants range from 75 to 900 MREM/HR, while dose rates at lower elevations are in the range of 15 to 250 MREM/HR.

Expected neutron dose rates for SHNPP at lower elevations

should, therefore, be about one seventh of those of the referenced operating plants and can thus range from 10 to 130 MREM/HR at intermediate floors and from 2 to 36 MREM/HR at lower floors.

Streaming gamma dose rates at the above referenced operating plants were measured to be one fourth or less of the neutron dose rates for general containment areas.

A similar ratio is expected for Shearon Harris.

These considerations indicate that the neutron streaming shields at Shearon Harris Nuclear Power Plant will provide sufficient shielding to achieve ALARA exposure to occupants in containment while the reactor is at power.

The anticipated frequency at which entries will be made into containment, the number of people making these entries, and their stay times are as follows:

~Fr e cene Number of Peo le

~Ste Time entry per day Two (2) Operators minimum when the plant is critical.

30 minutes minimum 2.

Health Ph sics - one (1)

Two (2) Technicians entry per week 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

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description of the cavity around reactor vessel, the primary piping, the missile shield, and the primary shield.

In order to determine the effectiveness of the shielding afforded on reducing the dose rates due to neutron streaming on operating deck, calculations are performed with and without the shield in place.

The effective factor of reduction for gamma and neutrons of the installed shield is 5.7.

A check on the accuracy of the modelled geometrical representation of the reactor cavity and its surroundings has been obtained by verification between the computer generated pictures of the model and general arrangement drawings at various elevations and sections.

The neutron histories start on the surface of the reactor pressure vessel with energies and directions determined by processing the angular fluxes at the outermost mesh of the vessel determined by the DOT-III (Reference 12.3.2-B) calculations with the DOMINO code (Reference 12.3.2-C),

which is explicitly set up to provide the proper source information for the MORSE Program.

Variation in source strength along the circumference has been neglected for conservatism.

The MORSE calculation has been performed in two stages utilizing a MORSE to MORSE coupling technique.

The first stage stops the random walk either at the vessel flange level or just above the neutron shield.

In the second stage particles which escape at the flange or above the neutron shield are recorded on tapes which are used as inputs to MORSE runs utilizing the entire three dimensional geometrical model.

No biasing technique has been used for either stage of the calculations.

It is known from experience at operating plants that the dose rates in containment are due primarily to fast neutrons.

The Monte Carlo calculations of the dose rates in containment have been limited to the energy groups of fast neutrons emerging from the reactor vessel with energies above 0.11 MeV.

However, in order to be conservative, the response to the energy groups considered is followed down to epithermal energies.

A thirty percent uncertainty is anticipated in using the Monte Carlo calculational techniques.

Table I lists the estimated neutron dose rates at selected locations within containment.

While the computed neutron dose rates on the containment operating floor are relatively high, the dose rates in general areas of lower elevations, where personnel may require access during normal operation, are expected to be much lower due to the shielding afforded by various internal components and structures.

The computed values of the dose rates at lower elevations are less reliable due to difficulty in modelling the problem, considering the more complex geometry, which causes the large uncertainty in the results.

A reasonable estimate of the expected dose rates can be made by comparing the dose rates measured at several operating plants at lower elevations with the corresponding measured dose rates at operating floors.

A ratio of these dose rates in conjunction with its predicted dose r ates at the operating floor is used to estimate the anticipated dose rates at lower elevations of SHNPP.

Measurements conducted at PMR Plants such as Calvert Cliffs (Reference 12.3.2-D), St. Lucie 1 (Reference 12.3.2-E) and Millstone 2

TABLE I NEUTRON STREAMING DOSE RATES IN CONTAINMENT DURING NORMAL OPERATION LOCATION 1)

Manipulator Crane Parked along Refueling Cavity El.

286.00')

Fuel Transfer System Control Panel El.

286.00'EUTRON DOSE RATE (REM/HR) 8.7 7.6 3)

Edge of Refueling Cavity near Steam Generator 1C - SN El.

286.00')

Missile Shield on Reactor Vessel Integrated Head El.286.00'1.913.5 5)

Emergency Escape Lock El.

261.00')

Grating Floor El. 261. 00'.

12 0.01

0. 13 7)

Personnel Lock El.

236.00')

Grating Floor El.236.00'.034 0.002 0.036

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Refer ences 12.3.2-A 12.3.2-B 12.3.2-C 12.3.2-D 12.3.2-E 12.3.2-F DLC 23, "CASK 40 Group Coupled Neutron and Gamma Ray Cross Section Data",

ORNL - RSIC, 1974.

N. R.

Rhoades and F.

R. Mynatt, "DOT-III, Two-Dimensional Discrete Ordinate Transport Code",

ORNL-TM-4280, 1973.

M. B. Emmett et al.,

"DOMINO":

A General Purpose Code for Coupling Discrete Ordinates and Monte Carlo Radiation Transport Calculations, ORNL-4853, 1973.

Baltimore Gas

& Electric Co. "Calvert Cliffs Unit No.

1 FSAR Amendment No. 51, Docket No. 50-317",

November -1973.

Florida Power b Light, "St. Lucie Unit 1 Docket No. 50-355 License Condition D, Neutron Streaming Shield", April 1977.

Northeast Nuclear Energy Company "Radiation Survey Results in and Around Millstone Unit 2 Containment Building," Docket No. 50-336, April 1976.

(6642PSA cfr)

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