ML20028E318
| ML20028E318 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 01/11/1983 |
| From: | Edelman M CLEVELAND ELECTRIC ILLUMINATING CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| 10CFR-050.55E, 10CFR-50.55E, RDC-46-(82), NUDOCS 8301210217 | |
| Download: ML20028E318 (3) | |
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P O Box 5000 - CLEVELAND, OHlo 44101 - TELEPHONE (216) 622-9800 - lLLUMINATING BLOG. - 55 PUBLIC SOUARE Serving The Best Location in the Nation MURRAY R. EDELMAN January 11, 1983 VICE NtESIDENT NUCMAR Mr. James G. Keppler Regional Administrator, Region III Of fice of Inspection and Enforcement U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137 RE: Perry Nuclear Power Plant Docket Nos. 50-440; 50-441 Stress Analysis for Centrol Rod Drive System [RDC 46(82)]
Dear Mr. Keppler:
This letter represents our final report pursuant to 10CFR50.55(e) concerning the design deficiency of the Control Rod Drive Hydraulic Piping System. This was first reported by Mr. R. Vondrasek of The Cleveland Electric Illuminating Company to Mr. J. E. Konklin of your office on February 9, 1982. Previous interim reports filed on this subject were dated March 11, 1982, and August 20, 1982.
This report includes a description of the deficiency, an analysis of safety implication, and corrective action taken to revise the design of the piping system.
Description of the Deficiency A design deficiency presently exists in the Control Rod Drive (CRD) system in that the CRD insert and withdrawal piping and associated piping supports were not designed to fully accommodate the influence of possible hydrodynamic loads due to operation of the scram valves which are located within the hydraulic control units. Relevant design information critical to the pipe stress analysis and pipe support design was neither supplied by the Nuclear Steam System ,
supplier, General Electric, nor requested by the Architect Engineer, Gilbert Associates, Inc., (GAI) prior to completion of the system piping design.
Subsequent to learning of the design transient, Gilbert Associates attempted to quantify the effects of the rapidly (20-60 milliseconds) opening inlet and exhaust scram valves on the piping design. Initially GAI performed a static analysis using dynamic load factors and found that certain areas of the CRD piping and piping support system exceeded the allowable stress limits defined by ASME Section III for Class 2 piping and component supports.
8301210217 830111 PDR ADOCK 05000440 s PDR
'J AN 1 4 083 TEA 7
. Mr. James G. Keppler January 11, 1983
- Analysis of Safety Implications Those portions of the Control Rod Drive System which are necessary to accomplish the safety function of inserting negative reactivity to shut down the reactor are classified as safety class 2. The scram event as mentioned in the description of the deficiency is analyzed as two cases. The first is the hot scram case in which the RPV is at full power and the insert and withdraw lines are maintained at reactor pressure when the scram occurs. The second is the cold -
scram case in which the RPV is .being brought up to full power. The pressure in the insert and withdraw line follows the reactor pressure up to full power.
For the cold scram case, the RPV is assumed to be 0 psi when the scram event occurs.
If insert and/or withdrawal piping would crimp due to plastic deformation during a scram event as might occur in areas of stress in excess of ASME III allowables, the CRD may become inoperable. As a result of CRD inoperability, the operator can initiate operation of the Standby Liquid Control System (SLCS) to shut down the reactor. In addition, the Redundant Reactivity Control System (RRCS) would recognize this condition as an ATWS event (Anticipated Transient Without Scram) and RRCS would begin to shut down the reactor. The RRCS is the logic to shut down and maintain the reaccor in a suberitical condition. These systems include Alternate Rod Insertion (ARI), Feedwater Runback, Reactor Recirculation Trip and Standby Licuid Control System (SLCS).
With regard to the safety implication of this deficiency to plant and public, the loss of functionability of the CRD system would not prevent the nuclear
- power plant from shutting down. The RRCS provides the same function as the Reactor Protection System (RPS) except that RRCS functions with the recognition that the CRD system is not initiated during a plant transient. The plant could attain cold shutdown even if the CRD system were totally impaired by deformation of the insert and/or withdrawal piping. Any radioactive release resulting from this condition would be contained within the plant and within 10CFR100 limits. Therefore, the health and safety of the public would not be affected.
Corrective Action Taken Although the above safety evaluation states that damage to the insert and withdrawal piping would not affect the public safety, it is clear that such damage might severely limit future plant operation. Therefore, steps have been taken by CAI to modify the CRD insert and withdrawal piping and piping support design to re-establish compliance with ASME,Section III stress 4
allowables for piping and component supports. These steps have been:
- 1. Modification of piping socket weld size and contour as necessary to reduce the number of locations of overstress. This was initially reported in our interim report to the NRC Office of Inspection and Enforcement dated March 11, 1982. This change has been implemented on joints not completed shortly af ter the deficiency was discovered.
, Mr. James G. Keppler January 11, 1983
- 2. In an attempt to refine the initial piping stress analysis and subsequent piping support design, CAI enlisted the assistance of a design consultant, ECHO Energy Consultants, of Oakland, California. ECHO assisted CAI in a comprehensive review of all CRD design transients. In addition, use of ECH0's Proprietary computer program IMPULSE-I to model these transients has resulted in reduction of unnecessary conservatism in the piping stress analysis and support design. Thus, the impact of final design revisions has been limited to pipe support modifications which may be implemented after installation of the previous design. These modifications are described below.
- a. In order to accommodate an overstress in the insert piping due to scram transients immediately inside drywell, GAI has determined that the redesign of an existing 2-way guide resulting in a 3-way guide will alleviate the overstress. This will be accomplished by the addition of pipe clamps which will provide axial restraint. Final design details are currently in process.
- b. Existing supports have been reviewed for scram transient eff acts and future modifications will be required to reduce support overstress (rather than pipe overstress). The supports which will require modification have been identified. The nature of the changes will not affect the piping as all involve modification at the structural attachment locations. Final design details are currently in process.
It is expected that final pipe support design details, as described in 2a and b above, will be available for implementation by April 1,1983.
Please call if there are any questions.
Sincerely, Murray . Edelman l Vice President l Nuclear Group t
MRE:pab i cc: Mr. M. L. Gildner i
NRC Site Office
! Director Office of Inspection and Enforcement l
U.S. Nuclear Regulatory Commission Washington, D.C. 20555 U.S. Nuclear Regulatory Commission I c/o Document Management Branch Washington, D.C. 20555
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