ML20197G092
| ML20197G092 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 11/21/1983 |
| From: | Knight J, Knight T Office of Nuclear Reactor Regulation |
| To: | Novak T Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-0684, CON-WNP-684 NUDOCS 8312070259 | |
| Download: ML20197G092 (8) | |
Text
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DE:MEB Read!rg File gy 7.1 33 t'EP. ORA!IDUi1 FOR: Thonas M. flovak, Assistant Director for Licensing, DL 1
FR0!1:
James P. Knight, Assistant Director for Components & Structures Engineering, DE
SUBJECT:
IflPUT TO WNP-2 SSER The t'echanical Engineering Branch has completed its review of several responses provided by the applicant subsequent to the issuance of the UNP-2 SER Supplement flo. 3 (April 1983). These itens include:
(1) the recuirements for snubber precperational testing, (2) the results of the dynamic systen analysis of reactor internals under faulted conditions, (3) design of component supports, and (4) inservice testing of purps and valves.-
Attached is our safety evaluation of the applicant's responses to be includea in the,, ext supplement to the UtlF-2 SER.
James P. Knight, Assistant Director for Comparents & Structures Engineering Divisien of Engineering Attachrent: As stated cc w/ attach.:
R. Vollner, DE R. Bosnak, DE A. Schwencer, DL-H. Bramer, DE R. Auluck, DL P. Chen, ETEC Y. Li, DE
Contact:
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3.9.2.1 Piping Preccerational Vibration and Dynamic Effects Testing In Section 3.9.2.1 of the SER, the staff identified a confirmatory item concerning the preoperational testing requirements for snubbers. We have reviewed the 1.nformation on preoperational testing requirements for snubbers described in Section 14.2.1.2.3.17 of Amendment 29 to the FSAR.
One area of concern during the review was the implementation of our requirement that if there is a lapse of six months between the initial snubber inspection / test and the system preoperational test, portions of the inspection requirements need to be repeated.
In a letter dated November 3,1983, G. C. Sorensen to A. Schwencer, the applicant stated that the Supply System has incorporated the six month reinspection requirement in System Line-Up Test, SLT 303.0, paragraph 4.2.3.2.
Based on a review of the information submitted'b'y the applicant, we have determined that the applicant's response satisfies the preoperational testing requirements described in the letter from R. L. Tedesco, NRC to the applicant, R. Ferguson dated March 16, 1981 and, therefore, the staff considers this confirmatory issue closed.
3.9.2.4 Dynamic System Analysis of Reactor Internals Under Faulted Conditions In Section 3.9.2.4 of the SER, the staff stated that the applicant committed to document the final results of its analysis of the reactor internals and unbroken loops of the RCPB, including the supports, for the combined loads due to a simultaneous LOCA and SSE including the
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. effects of annulus pressurization. We have reviewed the results of the applicant's analysis provided in Section 3.9.2.4 of Amendment 29 to FSAR. We find that the applicant's results meet the applicable design basis acceptance criteria described in FSAR Table 3.9-2b and, therefore, the staff considers this confirmatory issue closed.
3.9.3.3 Comoonent Supoorts In Section 3.9.3.3 of the SER, the staff stated that the applicant is committed to assess its component support design adequacy. The staff's specific concern was that for the design of component supports, the stresses produced by seismic anchor point motion of piping and the thermal expansion of piping should be categorized as primary stresses.
Expansion stresses in the support themselves may be categorized as secondary stresses.
In References 1, 2 and 3, the applicant has performed an assessment to evaluate its component support design adequacy with respect to the staff's position. As a result of its assessment, the applicant concluded that its component support design either explicitly did include the loads due to seismic anchor point motion anc thermal expansion of piping as primary stro Jes, or there was sufficient design margin to accommodate them.
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To demonstrate the actual design margin, five representative anchor groups (122 supports) were reanalyzed by the applicant using the inherent design conservatisms, i.e., the new seismic spectra resulting
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. from the finite element model (as permitted by Standard Review Plan 3.7.2) and increased damping values (0.5% damping vs. 2%/3% per Regulatory Guide 1.61). When these revised design inputs were applied, the average load reduction was about 20%. Furthermore, the applicant indicated that all the existing support hardware was found capable of carrying these refined loads by either inspection or reanalysis of the support hardware.
In Reference 4, a detailed analysis of an additional anchor group (32 supports) was performed by the applicant to further substantiate its previous conclusion. Based on a review of the information in References 1, 2, 3 and 4, the staff has determined that the applicant's component Support design adequacy assessment has demonstrated compliance with the staff position and, therefore, the staff considers this confirmatory issue closed.
3.9.6 Inservice Testing of Pumos and Valves In Section 3.9.6 of the SER, the staff identified a confirmatory issue regarding the inservice testing of pumps and valves which depended on the applicant's submission of the required system drawings. The applicant has submitted these required system drawings and we have reviewed the information on the pump and valve inservice test program d scribed ln the applicant's submittal (Reference 5).
In Reference 5, the applicant has requested relief from the requirements of Section XI of the ASME Code.
Based on the information 1-the applicant's inservice testing
. program and system drawings, we find that it is impractical for the applicant to meet some of the requirements of the Code. Therefore, pursuant to 10CFR Part 50.55a, the relief that the applicant has requested from the pump and valve testing requirements of 10CFR Part 50, Section 50.55(g)(2) and (g)(4)(1) is granted for a period of the initial 120 month period during which we complete our detailed review.
Therefore, the staff considers this confirmatory issue closed.
If, as a result of our detailed review, the staff identifies additional requirements for the inservice testing of pumps and valves, we will require that the applicant comply with them.
The staff has also addressed the issue of periodic leak testing of pressure isolation valves. These valves are placed in series to form the interface between the high pressure reactor coolant system and several low pressure systems. The leaktight integrity of these valves l
must be ensured by periodic leak testing to prevent exceeding the design pressure of the low pressure systems. The staff has reviewed the information of the Section 3.4.3.2 of WNP-2 Technical Specification l
l pertaining to this issue. The pressure isolation valves to be tested are listed in the Technical Specifications.
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The applicant has agreed to categorize their pressure isolation valves for the low pressure /high pressure core spray, residual heat removal, and reactor coolant isolation cooling systems, as Category A or AC.
l These categorizations meet our requirements and we find them acceptable.
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i Pressure isolation valves are required to be Category A or AC and to meet the appropriate valve leak rate test requirements of IWV-3420 of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, and demonstrate an allowable leakage rate not in excess of 1 gallon per minute for each valve as stated in the Technical Specifications. The applicant has committed to test all pressure isolation valves to this criteria.
The leak testing should be performed following each valve disturbance, when coming down to cold shutdown during an outage, or during restart.
Leak testing should be performed prior to entering mode 2 during restart.
Leak testing at this time assures that the pressure isolation valves are performing their intended function prior to entering power operation. The time during shutdown for leak testing is specified in the Technical Specifications.
Limiting Conditions for Operation in the Technical Specifications, requires corrective action, i.e., shutdown or system isolation when the leakage limits are not met. Also, surveillance requirements, which will state the acceptable leak rate testing frequency, is provided in the Technical Specifications.
We conclude that the applicant's commitments to periodic leak testing of pressure isolation valves between the reactor coolant system and i
low-pressure systems will provide reasonable assurance that the design pressure of the low-pressure systems will not be exceeded, and is acceptable. Criterion 55 of the General Design Criteria of Appendix A of 10 CFR Part 50 addresses.this issue in part.
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References 1.
Letter from G. D. Bouchey to A. Schwencer, " Nuclear Project No. 2 Confirmatory Issue #7 - Component Supports", dated December 30, 1982 2.
Letter from G. D. Bouchy to A. Schwencer, " Nuclear Project No. 2 Safety Evaluation Report (NUREG-0892) Confirmatory Issue No. 7 -
Component Supports", dated June 30, 1983 3.
Letter from G. C. Sorensen to A. Schwencer, " Nuclear Project No. 2 Safety Evaluation Report (NUREG-0892) Confirmatory Issue No. 7 -
Component Supports", dated September 14, 1983 4.
Letter from G. C. Sorensen to A. Schwencer, " Nuclear Project No. 2 Supply System Response to Revised FSAR Question 110.42", dated October 28, 1983 5.
Letter from G. C. Sorensen to A. Schwencer, " Nuclear Project No. 2 Pump and Valve Inservice Test Program Plan Update (Revision 2)",
dated November 4, 1983