ML20210T241
| ML20210T241 | |
| Person / Time | |
|---|---|
| Site: | Satsop |
| Issue date: | 08/28/1975 |
| From: | Maccary R Office of Nuclear Reactor Regulation |
| To: | Deyoung R Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-1682 NUDOCS 8605290807 | |
| Download: ML20210T241 (6) | |
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1 AUG 2 81975 Dockat Nos.: 5 08 6 509 MS: 24-1 R. C. DeYeung Assistaat Director for Light Water Reesters, Group 1 Division of Reester Licensing WASRINGT001 FURLIC POWER SUFFLY SYSTEM, DOCERT NOS. 50-508 ti 509 Flaat Name: Washington Nuclear Project No. 3 (UNF-3) and No. 5 (WNF-5)
Decket Mos.: 50-508 and 509 Licensing Stage: PSAR Responsible Branch and Project Manager: LWR l-3, P. O'Reilly Responsible Branch and Technical Reviewers: MEB, F. Cherny, P. Y. Chen Requested Completion Date: 8/29/75 Description of Response: Safety Evaluation Report Sections 3.6 and 3,9.2.5 Review Status: Complete The PSAR subsitted by theaapplicant through Amendment 21 has been reviewed by the Mechanical Engineering Branch, Division of Technical Review.
The Mechanical Engineering Branch area of review concerns the design critoira of Sections 3.6, 3.9, 3.10, 4.2, 5.2 and 5.5 of the Standard Format (Regulatory Guide 1.70) dated October, 1972. Since the WP-3 and 5 PSAR references CESSAR, only non-CESSAR portions of these sectbus have been reviewed. Accordingly, only non-CESSAR sections of the SER have been prepared.
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Sections 3.9.1.1, 3.9.1.2, a portion of 3.9.2, 3.10 and 5.2.8.7 were l
attached to our letter of June 16, 1975.
I Section 3.6 and a portion of Section 3.9.2 (3.9.2.5) are attached to this letter. For agreement with P. O'Reilly these sections include the following open items which we believe, based upon verbal discussions with the applicant, will be resolved in future amendments to the PSAR and/or via topical reports to be submitted. The eyes itses are the folleeing:
Secties 3.6 (1) The applicant has referessed a topical report, RTR-1002 entitled " Design Considerations for the Protestion from the Effect of Pipe Emptanre." This report has met been formally submitted to IRC. Based spes verbal dis-l emesiens with the applisant, we are ender the 1syression that the report I
has met been esepleted, se reenement that a firm date be obtained from the applisant for submittal of this tepisal.
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R. C.DeYoung (2) In Section 3.6.4.2 the jet impingement analysis for Case E is not adequate. The applicant has verbally agreed to revise the analysis method for the case. We expect the revision will be included in a future amendment.
Section 3.9 In Section 3.9.2.5 ref erence e,s made to two recettcly developed computer codes, LOADFACT and Pipestress 2010. The app 41 cant has stated in the PSAR that these codes will be utilized to analyze for Cae effects of dynamic loads resulting from opening of safety and relief 4]ves. The applicant has not verified in the PSAR the validity of utilizing these codes fe: thr.se types of analyses. Based upon verbal discussicus with the applicant, we are under the impression that the applicability of these codes will be verified in the topical report ETR-1002 ref erenced under Sect, ton 3.6 above. A firm date for submittal of this report should be obrained from the applicant.
In addition to the above open items, we would like to note the following which have not been specifically inserted in our input to the WPPSS 3 and 5 GER.
Section 3.9.1.3 In amendment 36 of CESSAR, the WPPSS 3 plant has been designated ao the CE System 80 prototype plant for satisfying the Regulatory Guide 1.20 ~ equire-r nents for a comprehensive vibration assessment program. Through Ame::deunt 1]
of.the WPPSS PSAR, the applicant has not adopted Amendment 36 of CESSAR. We assame that the applicant will adopt this CESSAR amendment in a future amend-ment to the WPPSS PSAR and accordingly, we have not included this item in our SER input.
Section 3.9.1.5 Also MEB is reviewing, on a generic basis with all PWR vendore, the effect on the design of the reactor vessel supports of a heretofore grossly underesticaSed post LOCA load which would act on these supports, This item hris beeA addressed in the CESSAR Report to ACRS. We assume that since WPPSS 3 and 5 are CESSAR standard plants the applicant will adopt whatever solution to this problem Combustion Engineering commits to in CESSAR. Accordingly, we have not aduressed this item specifically in our WPPSS 3 ani 5 SER input.
a a.e R. R. Maccary, Assistant Direct r for Engineering Division of Technical Review cc w/ encl:
R. E. Heineman, TR F. C. therny, TR
R. J. Bosnak, Tn R. L. T,oyd, RL l
H. L Brnmer, TR W. G. Mcdonald, MIPC
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MECHANICAL ENGINEERING BRANCH DIVISION OF TECHNICAL REVIEW
's WASHINGTON NUCLEAR PROJECT NO. 3 & N0. $
JJ SAFETY EVALUATION REPORT
~3.6 frateation Against Dynamic Effectu Associatrd with the Postul5ted Rupture 'of Figing With respect ce piping systems locate 3 inside contui.nment, the applicant states that the criteria to be Caployed for detet.wination of b
the systems which are evaluated, the locations and typea of piping 1.runks
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which are postulated, and the prottatioi; measures against pipb whip to be 4
provided, will be consistent with the. acceptable criteria of Regulatoyy Guide 1.46, " Protection Against Pipe Whip laside Containment."
The methods uecd for forculating the hydro-dynamic forcing funct$ons induced by postulsted pipp rupture and the dynamic analyst.? for the resultir3 pipe whip motion provide an acceptable basis for restraint design. The use of thesc methods will provide reasonable assurance that the cantainment structure, unaffected system compancats, and those systems impor.taat to safety which are in close pteximity to the systems in which postulated pips failures are assumed to occur, pill be protected. These provisions for protection against che dyr.smic ef fectp associated with pipe ruptures and the retulting dicchargin6 coolant, provide reasonable ars: trance that, in the event of the tambinLd leadings imposed by an earthquaic of the pagnitude of ths safe. stutdown earthqeake l
and a concurrent single pipe brer.k at one of the design basis break
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W. 4 locations, the following conditions and safety functions will be accommodated and assuted:
(a) the magnitude of a design basis loss-of-coolant accident cannot be aggravated by potentially multiple failures of piping, and (b) the reactor emergency core cooling systems can be expected to perform their intended function.
The criteria used for the identification, design, and analysis of piping systems inside containment whcre postulated breaks may occur constitute an acceptable design basis in meeting the applicable requirements of Criteria 1, 2, 4, 14, 15, 31 and 32 of the NRC General Design Criteria.
With respect to piping systems located outside containment, the
' applicant will designate design basis break locations throughout all high energy piping systems. These postulated break locations will be chosen on the basis of highest relative stress, or significant changes in flexibility of the piping. The protection to be provided against the dynamic effects of postulated pipe breaks and discharging fluids in piping systems containing high energy fluids and located outside the containment is adequate to prevent damage to structures, systems and components to the extent considered necessary to assure the maintenance of their structural integrity. Such protection will provide reasonable assurance that the safe snutdown of the reactor can be accomplished and maintained as needed.
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-t In addition, for those piping systems not considered as high energy systems, the applicant will postulate sufficient Icakage cracks to assure that essential equipment and comoonents are protectw1 from fluid spraying, floodicg, and consequent environmactal conditione deieloped. This is consistent with criteria
- set forth in U. S. Nuclear Regulatory Comoission Standsrd Review Plan 3.6.2.
.The applicant will submit a topical report, ETR-1002- % sign Consideration for the Protection from the Effect of Pipe Rupture", to define in more detail criteria which will be used in providing protect.icn from these postu2ated events. The report will also summarize the ana3ytical methods which describe the immediate consequence of the postulated failure. We vill require _ that the criteria described in this topical report be consistent with that set forth in Standard Review Plan 3.6.2 ref erenced above.
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i t 3.9.2.5 Design and Installation Criteria, Pressure Relieving Devices The criteria used in developing the design and mounting of ASME 2 and 3 safety and relief valves provides adequate assurance that, under discharging conditions, the resulting stresses are expected not to exceed the allowable design stress and strain limits for the materials of construction. The applicant has referenced two recent]y developed computer codes, LOADFACT and Pipestress 2010 for use in analyzing the effects of dynamic loadings associated with the sudden operation of pressure relieving devices. The app 3 f cant has stated that topical reports will be submitted to NRC which will establish the validity of these computer codes for this type of analysis. We will require that these validations comply with criteria set forth in items II.2 ef U. S. Nuclear Regulatory Commission Standard _ Review Plan 3.9.1.
LimitJng the ctresses under the loading combinations associated with the act.uation of the pressura relief devices provides a conservative basis for the design of the system components to withstand these loads without loss of structurel integrity and impairment of the overpressure protection function.
The criteria used for the design and the installation of ASME Class 2 and 3 overpressure relief devices constitute an acceptable design basis in meeting the applicable requirements of NRC General Design Criteria 1, 2,,4, 14 and 15 and are consistent with those specified in Regulatory Guide 1.67, " Installation of Overpressure Protection Devices."
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