ML19256A523

Summary of 781221 Meeting Re Neutron Shield Design.Applicant Will Forward Qualifications of Design by mid-January 1979 & Request a Safety Evaluation by 790331

ML19256A523
Person / Time
Site:	Millstone
Issue date:	12/21/1978
From:	Conner E Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
TAC-46174, NUDOCS 7901080333
Download: ML19256A523 (10)

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NUCLEAR REGULATORY COMMiss10N

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WASHINGTON, D. C. 20666 December 21, 1978

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Docket No.: 50-336 LICENSEE: NORTHEAST NUCLEAR ENERGY COMPANY (NNECO)

FACILITY: MILLSTONE UNIT NO. 2

SUBJECT:

SU!HARY OF MEETING HELD ON NOVEMBER 29, 1978 REGARDING NEUTRON SHIELD DESIGN Introduction On November 29, 1978, representatives of NNECO met with representatives of the staff at Bethesda, Maryland. A listing of the attendees is presented as Enclosure 1.

The meeting had been requested by NNECO to discuss their revised neutron shield design for Millstone 2.

Background

In a meeting on September 28, 1976, NNECO sumarized the neutron streaming problem at Millstone 2 and discussed possible resolutions.

By letter dated April 15, 1977, NNECO documented their intention to respond.to our letter of February 10, 1977 and install a neutron shield during the refueling outage following Cycle 2 operation. The proposed shield design was similar to that proposed by NNECO letter dated Decenter 4,1975.

By letter dated June 6,1978, NNECO presented their design of a permanent neutron shield made of barated silicon rubber. This de-sign has been abandoned, because of high cost and excessive personnel radiation exposure during installation, in favor of a segmented water tank design as presented in the NNECO letter of November 13, 1978.

Discussion NNECO provided the Enclosure 2 agenda for the meeting.

In the intro-duction, Mr. Kacich presented the history of the neutron shield review for Millstone 2. ' This review is in response to license condition 2E.

He pointed out that the radiation exposure: problem was only inside the containment. He then provided a sumary of the review since our last meeting in September 1976. NNECO plans to install the shield as presented in their November 13, 1978 submittal during the Cycle 3 reload outage scheduled for the spring of 1979.

7901086333

Mtg. Sumary for NNEC on 11/29/78 In discussing the anticipated shielding design, NNECO (Kacich) passed out Enclosure 3 and reviewed each item. The staff (Block) questioned the neutron spectrum to be used in the analyses. Mr. Kacich said the spectrum would be detemined by activation techniques. Dr.

Nehemias questioned the water level monitoring to be used. Mr. Weyland stated that the shield tank is sectionalized into 16 compartments and will be leak tested prior to use. An increase in the containment radiation level will pmvide indication of shield tank leakage. Mr.

Kacich said they should need to move the shield for refueling and possibly incore detector work, only. In response to Mr. Block's question on gama dose of shield during storage, Mr. Kacich said the shield storage would be on top of the steam generator enclosure away from the general work area. Mr. Block questioned the type of personnel monitoring used at Millstone 2.

Mr. Kacich responded that they use the stay time method for monitoring.

In a discussion of the materials of construction, Mr. Kacich said the actual material had not been selected yet.

Mr. Whittlessy passed out the Enclosure 4 sumary of shield design and analytical qualification. He discussed the forced convection' cooling design necessary to keep the shield water below 200 F.

He also discussed the burst grooves machined to between 30 and 40 mits depth in the tank bottom and lid. Mr. Zudans questioned '.he burst panel qualification. Mr. Whittlesey said testing was underway but qualification would be by analyses methods. Mr. Shum questioned the thermal insulation movement during tank rupture following a LOCA.

Messrs. Kccich and Weyland responded that the analyses to be done would use'd double-ended-break of cold leg.

We then discussed the qualification figure in Enclosure 4 and the seismic design. Mr. Zudans cautioned NNECO to use the dampening values reviewed in the FSAR. Mr. Zudans suggested that the movement path be reviewed and controlled to prevent damage of any components if the shield is dropped during movement.

It was further recomended that material stresses be tabulated in the submittal similar to the data presentation for a spent fuel pool modification request.

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Mr. Kacich said the submittal with all qualifications should be ready in mid-January 1979. They request that our Safety Evaluation be issued by the end of March 1979.

Eben L. Conner, Project Manager Operating Reactors Branch #4 Division of Operating Reactors

Enclosures:

1.

List of Attendees 2.

Meeting Agenda 3.

Items on Shielding 4.

Summary of Reactor Cavity Shield Design & Analytical Qualification

ENCLOSURE 1 LIST OF ATTENDEES MEETING WITH NNECO ON 11/29/78 REGARDING NEUTRON SHIELD DESIGN MEETING NRC E. Conner F. Harper J. Nehemias D. Shum J. Zudans S. Block E. Lantz NUSCO

3. Weyland R. Kacich EDS Nuclear M. Whittlesey BG&E R. Olson 4

ENCLOSURE 2

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MILLSTONE WIT NO. 2 NEUTRCN SHIELDING MEETING AGENDA NOVEMBER 29, 1978 (1) INTRODUCTION, BACKGROLHD RICK KACICH (2) MTICIPATED SHIELDING RICK KACICH (3) GENERAL DESCRIPTION MD DESIGi BASIS CONSIDERATIONS MARK WHITTLESEY (4) INSTALLATION MD REMOVAL STEVE WEYLMD (5) QUESTIONS MD MSWERS

ENCLOSURE 3 MILLSTONE UNIT 2 SHIELDING Shield geometry modeled into DOT 3.5 Computer Code Source term based on neutron fluence measurements taken at flange elevation in the annular gap.

MLdel assumes 3.5 inch gap between the inner edge of the shield and the closure head.

Model assumes 3.5 inch gap between refueling pool floor and shield bottons. Overlap of 4 inches (i.e. radially outward streaming path).

Neutron fluence assumed to be isotropic in the upward direction except for directly under the reactor flange.

Under reactor flange, i.e. in the 3.5 inch air gap, fluence assumed to be in a vertically upward direction, perpendicular to the flange.

Tank depth is 18 inches, minimum water height is 16 inches.

With the above modeling assumptions, calculations indicate that no significant increase in attenuation is achieved after the first 14 inches.

This design yields an attensation of a factor of forty, with attenuation f actor defined as the ratio of aver. age dose rate at the flange level with no shield in place to tre average dose rate with the shield in place.

Anticipate that less than 5 man-rem will be espended dur'.ng initial installation, with subsequent installations and removals involving less than 2 man-rem.

ENCLOSURE 4

SUMMARY

OF SULLSTONE UNIT 2 REACTOR CAVITY SHIELD DESIGN AND ANALYTICAL QUALIFICATION November 29,1978

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Qualification Design Criteria:

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Qualification of Design Groove el and tanks?

• Additional major design considerations include leakage, installation and removal (ALARA), and heat flow.

MEETING

SUMMARY

DISTRIBUTION ORB #4 Mr. W. G. Counsil, Vice President Nuclear Engineering & Operations Northeast Nuclear Energy Company P. O. Box 270 Hartford, Connecticut 06101 Docket File N

V. Noonan PDR P. Check ORB 44 Rdg G. Lainas NRR Rdg G. Knighton H. Denton Project Manager E. G. Case OELD V. Stello OI&E(3)

B. Grimes R. Ingram T. Carter R. Fraley, ACRS(16)

D. Eisenhut TERA A. Schwencer J. Buchanan D. Ziemann Meeting Summary File T. Ippolito Program Support Branch R. Reid NRC Participants F. Harper J. Nehemias D. Shum J. Zudans S. Block E. Lantz

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