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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION R_ ELATING TO MARK I CONTAINMENT PROGRAM - VACUUM BREAKER INTEGRITY BOSTON EDISON COMPANY - PILGRIM NUCLEAR POWER STATION 00CKET NO.: 50-293 1.
INTRODUCTION In addition to the evaluation of the supp ession chamber, torus attached piping, pressure relieving lines, etc, under the newly defined loadings, the Mark I containaent program required the assurance of the structural integrity l
of vacuum breakers during operation in all Mark I plants.
This additional requirement was categorized as a separate effort, as the adequacy of other components was already discussed in a separate Safety Evaluttion.
The Franklin Research Center (FRC) has performed an evaluation of the structural integrity of vacuum breakers in the Pilgrim Nuc*. ear Power Station (Pilgrim) for the NRC staff.
Results of the review are ',eported in the attached document, TER-C5506-328, "Structural Evaluation of the Vacuum Breakers (Mark I Containment Program), Pilgrim Nuclear Power Station."
FRC has concluded that actions taken by the licensee are adequate to restore the original design margin of safety for its vacuum breakers under the revised loadings in the Mark I containment.
NRC staff reviewed the attached document and concurred with the FRC findings.
II.
DISCUS $10N In the Pilgrim Mark 1 Containment there are ten 18" internal type vacuum breakers made t*y General Precision Engineering.
One Vactum breaker is mounted on the ient line end caps at the intersection of each vent line and the ring header in the suppression chamber.
Loadinga on Mark I structures and vacuum breakers are based on the General Electric Ccepany Kaport, NE00-21888, "Mark I Containment Program Load Definition Report," Revision 2, i
dated November, 1981.
For vacuum breakers, the loadings included are gravity, seismic, and hydrodynamic loads.
The hydrodynamic forcing functions were developed by Continuum Dynamics, Inc. by using a dynamic model of a Mark I i
pressure suppression system and the full scale test facility data.
The system model was capable of predicting pressure transients at specific locations in the v*nt system.
Loading across the vacuum breaker disc caused by pressure differentials based on test data was thus quantified as a function of time.
i This issue was reviewed and approved by NRC on December 24, 1984.
Loadings were combined according to the FSAR commitments.
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70 determine the structural integrity of the vacuum breakers, results from a finite element model and ANSYS program analysts were compared with design limits specified in tt,e ASME Boiler and Pressure Vessel Code,Section III, Civision 1, Subsection NC,1977 Edition and addenda up to Sumer 1977.
It was found that the pallet, the hinge arm, the hinge shaft and the hinge arm stud could become overstressed.
The licensee decided to remedy the situation by using different materials for these parts to increase their allowable stress limits.
By changing SA-516 Gr 70 (pallet and hinge arm) and SA-320 88 (hinge s5 aft and hinge arm stud) to SA-705 Gr630 (pallet) and SA-564 Gr 630 (hinge arm, hingF), the allowab e stress limits changed from 35 and 30 ksi, to 70 ksi.e sha at 1,100 Proper safety margins were thus restored.
III.
CONCLUSION The licensee has restored the safety margins of the Pilgrim vacuum breakers by replacing critical parts with adequate materials.
The corrective action is acceptable.
The staff therefore recomends that this issue be closed.
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Docket No.:
50-293 1
NEMORANDUM FOR:
Prbesult, Project Manager Project Directorett 81*
I 01visten of WR Licensing FROM:
Gus C. Lainas. Assistant Director Division of BWR Lictnsing l
SUBJECT:
SAFETY EVALUATION REPORT FOR GENERIC LETTER 83-28, I
liEM 2.1 (PART 1) (EQUIPMENT CLASS!FICATION)
(SRP SECTION 7.2, 17.2) FOR P!LGRIM NUCLEAR P0.iER ST U (04 Plant Name:
Pilgrim Nuclear Power Statitin l
Utility:
Rostnn Edison Company Docket No.:
50 293 TAC No.:
5?867 l
Licensing Status:
OR Resp. Directorate:
PAD 81/0PLA Project Manager:
P. Leech l
Review Branch:
PAE!/DPA l
Review Statut:
Incomplete i
The licensee was r+taired by Gererb f.ettia 13-?B Ittr ' ' f oo t.1) to cor:f irr-l that all components whose functioning is required to trip the reactor are identified as safety-related on all plant documentation and in information handling Systens that are used to control all activities perfonned on this safety-related equipment.
The licensee has responded and our review of the responses as docue nted in the enclosed contractor's report (EGAG-NTA-7188) l finds the licensee's responses to Generic Letter 83-28 Item 2.1 (Part 1) to be acceptable.
The enclosed SER documents our concurrence with the contractor's findings and also finds the licensee's responses for this item to be acceptable. We therefore consider Part 1 o' Item ?.1 to be closed by this action. sal.P input for tne review of iten 2.1 (Part 1) of Generic Letter i
83-28 is enclosed.
1 Generic Letter 83-28 Item 2.1 and its associated TAC number rensin incomplete because the Vendor Interface portion (Part 2) of this item has not been resolved.
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Enclosures:
dus C.
ainas. Assistant Director As stat?d Division of BWR Licensing cc:
R. W. Houston T. Novak C. E. Rossi J. Zwolinski M. Srinivasan
Contact:
D. Lasher, E!CSB/nPA X27200 J
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