ML20198S613

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Partially Deleted Rev 9 to Edcr 93-303, Vessel Internals Segmentation Component Removal Program 1995
ML20198S613
Person / Time
Site: Yankee Rowe
Issue date: 12/31/1993
From:
YANKEE ATOMIC ELECTRIC CO.
To:
Shared Package
ML20198S170 List:
References
FOIA-98-175 93-303, 93-303-R09, 93-303-R9, NUDOCS 9901110371
Download: ML20198S613 (39)
v • d • e


Text

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[p Cognizant Engineer c:

PORC Subcommittee PORC Committee P Centilli P Sheldon W Jones B Smith i

l

MEMORANDUM l

TO:

PORC August 10, 1993 l

FROM:

E.F.

Begiebing

SUBJECT:

Subcommittee Review of EDCR 93-303

-----===.__========================_===:r_===n==============

DISCUSSION:

A PORC Subcolitmittee meeting was held on August 9, 1993 to review the.following document:

.g3 EDCR 93-303 "Reactor Vessel Internals Segmentation Component Removal Program 1993" ATTENDEES:

E.F.

Begiebing - Chairman C.L.

Child - Chairman B.A.

Darcy R. Williams N.

Johnson S.M.

Litchfield D.

Calsyn B.

Cox N.

Fetherston i.

l Not in attendance but also submitted comments were:

- J. Kay

?..'May A' review of'EDCR 93-303 was conducted for technical content, compliance with Plant Technical Specifications and plant procedures.

The subcomnittee review resulted in the following comments:

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. CONCLUSION Based on the satisfactory resolution of the above comments, the i

~ PORC Subcommittee recommends approval of EDCR 93-303 " Reactor Vessel Internals Segmentation Component Removal Program 1993."

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p h.y M INPUT TO INSPECTION REPORT 50-29/93-05 i

Steam Generator Removal Prooram - 10 CFR 50.59 Safety Evaluations i

The purpose of this inspection was to review the licensee's 10 CFR 50.59 safety evaluations supporting the steam generator removal program.

i The j

inspection was performed in two phases; first, the licensee's three procedures j

covering the preparation of safety evaluations were reviewed. These i

procedures are listed and evaluated in the following paragraphs.

Second, the i

safety evaluations specifically related to the steam generator removal program were reviewed; these safety evaluations are contained in the Yankee Rowe Engineering Design Change Request (EDCR)93-302 entitled, " Steam Generators and Pressurize,r Removal - Component Removal Project 1993."

i Yankee Rowe Procedure No. AP-0001, Revision No. 25, " Plant Procedures" dated j

Ju 1993, is the top level document in the aration of safety evaluations.

i 25 l

The next level of procedure is AP-0200, Revision No.13, " Plant Modifications" dated July 1993. This document properly establiyhes controls and guidance for plant modifications.

1' The licensee's assurance of proper use of 10 CFR 50.59 is provided by i

j Procedure No. AP-0059, Revision No. 3, " Safety Evaluations" dated July 1993.

AP-0059 provides guidance for the preparation of safety evaluations related to 1

t modifications now that the lant is rmanently shut down and defueled.

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In addition to the basic requirements of 10 CFR 50.59, AP-0059 also contains guidance on meeting the criteria of the Commission's Staff Requirements Memorandum (SRM) of January 14, 1993. This procedure references the Generic Environmental i

{

Impact Statement (GEIS), NUREG-0586 and will serve to keep the environmental j

impacts of this program within the envelope of the GEIS analyses.

The inspectors reviewed an additional guidance document, Yankee Project Procedure (YPP) 32, Revision 0, " Guidance for Preparing a Safety Evaluation" dated June i

6,1993, that provides guidance for decommissioning activities that are e'

rformed rior to NRC ap roval of the Decommissioning Plan.

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i The EDCR contains the safety evaluations for steam generator and pressurizer removal, design criteria, a description of the program, calculations and drawings. Only the safety evaluations for the steam generator removal were reviewed in this inspection.

The EDCR was approved by the Plant Operating Review Committee on July 16, 1993.

Yankee's analysis demonstrates that there is no increase in the consequences or probability of any previously evaluated accident due to the steam generator removal program. The inspectors agree with this assessment as the only o

remaining design basis accident, since the plant is permanently shut down, is the fuel drop accident and we determined that the fuel puol, its piping systems and surrounding structures are sufficiently separated from this work m

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as to be unaffected by the work.

In addition, the inspectors verified that proper isolation of pool cooling systems and component cooling systems were considered in the safety evaluation.

Yankee evaluated a range of unanalyzed events.

The licensee found that the drop of a steam generator (SC) from a maximum height of 90 feet would represent the worse case accident not previously evaluated.

Yankee estimated that 7% of the solid radioactive material inside a dropped SG would have to be released and transported to the site boundary before the design basis dose criteria would be exceeded. These criteria are the EPA Protective Action Guidelines (PAGs) which are more conservative than 10 CFR Part 100 limits.

The licensee has adopted a defence-in-depth approach in order to prevent such a drop from occurring and mitigation devices should the drop occur.

The prevention feafures consist of: crane load testing in excess of the crane 150 ton capacity (a SG weighs 104 tons), careful selection of load path, QA and inspection programs for crane components, proposed use of experienced crane operators, use of senior supervisors in lifting operations and careful review of new and existing crane lift procedure documents. The mitigation features consist of a load transfer device and an impact limiting pad.

The load transfer device is designed to protect vulnerable features of the SG such as nozzles, protruding manways and temporary cover plates from impacts by routing the impact loads away from these vulnerable features. The impact limiting pad is designed to absorb the energy of a 90 foot dr'op of a SG through crushing of the pad. The 10 foot thick pad is made of dense foam.

The inspectors reviewed the unanalyzed events listed in the EDCR and agree that the 90 foot drop event is the worse case accident in this program.

Furthermore, the inspectors find that the prevention and mitigation approach of the licensee should serve to maintain any release from a SG drop below the EPA PAGs.

The licensee stated in ti.c EDCR that no margin of safety, as defined in the basis for any Technical Specification, is reduced by the steam generator removal activities. The inspectors verified this assessment and found the conclusion acceptable.

In addition to the safety evaluations performed by the licensee pursuant to 10 CFR 50.59, Yankee included in the EDCR an evaluation related to their conformance with the criteria of the SRM.

In summary, these evaluations resulted in the findings that these activities do not: foreclose release of the site for unrestricted access, significantly increase the cost of decommissioning, cause any significant environmental impacts not previously evaluated nor violate the terms of the existing facility license. The NRC staff in letters to the licensee dated April 16 and July 15, 1993 had stated that the NRC had no objection to the licensee's component removai program prior to approval of the Decommissioning Plan based on Yankee meeting these criteria.

The inspectors verified that the licensee in the EDCR properly evaluated these criteria.

The inspectors conclude that those portions of the component removal program safet/ evaluations related to the steam generator removal are acceptable.

7 2 2-93

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