ML20212F109
| ML20212F109 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 12/22/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20212F113 | List: |
| References | |
| TAC-08731, TAC-08732, TAC-80731, TAC-8731, TAC-8732, NUDOCS 8701090375 | |
| Download: ML20212F109 (10) | |
Text
.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO THE ELIMINATION OF LARGE PRIMARY LOOP RUPTURES AS A DESIGN SASIS NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNITS 1 AND 2 COCKET N05. 50-282 & 50-306 BACKGROUND.
By letters dated October 24, 1984, October 21, and November 5, 1985, Northern States Power Company (the licensee) provided technical information and requested an exemption to allow the application of " leak-before-break" technology as a basis for the elimination of protective devices installed to mitigate the dynamic effects resulting from postulated ruptures of Prairic Island Units 1 and ? primary coolant loops. The licensee submitted Westinghouse reports WCAP-10639, WCAP-10929, and WCAP-10931 as technical bases for the request.
By letter dated September 10, 1986, the licensee submitted Revision 1 of Westinghouse reports WCAP-10929 and WCAP-10931 in response to staff concerns on Prairie Island Unit 2.
The submittals were made in support of a request for an exemption to General Design Criterion (GDC) 4 in regard to the need for protection against dynamic effects from postulated primary loop pipe breaks.
On April 11, 1986, a final rule was published (51 FR 12502), effective May 12, 1986, amending 10 CFR Part 50, Appendix A, GDC 4.
The revised GDC 4 allows the use of analyses to eliminate from the design basis the dynamic effects of postulated pipe ruptures of primary coolant loop piping in pressurized water reactors.
In the " summary" section of the final rule, it is stated that the new technology reflects an engineering advance which allows simultaneously an increase in safety, reduced worker radiation exposures and lower construction and maintenance costs.
Irrplementation permits the removal of pipe whip 8701090375 861222 DR ADOCK 05000282 PDR
. restraints and jet impingement barriers as well as other related changes in operating plants, plants under construction and future plant designs.
Containment design, emergency core cooling and environmental qualification requirements are not influenced by this modification.
In the " supplementary ir. formation" section of the final rule, it is stated that acceptable technical procedures and criteria are defined in NUREG-1061, Volume 3, dated November 1984 and entitled " Report of the U.S. Nuclear Re5ulatory Commission Piping Review Committee, Evaluation of Potential for Pipe Breaks."
With the revised GDC 4, the exemption originally requested is no longer necessary. Using the criteria in'NUREG-1061, Volume 3, the staff has reviewed and evaluated the licensee's submittals and this report provides the staff's findings.
PRAIRIE ISLAND PRIMARY COOLANT SYSTEMS The primary coolant systems of Prairie Island Units 1 and 2 have two main loops each comprising a 34.6 inch diameter (outside) hot leg, a 36.9 inch diameter crossover leg and 32.8 inch diameter cold-leg piping. The materials for the primary loop piping are wrought stainless steel (376-TP316) and cast stainless steel (SA351-CF8M) for Prairie Island Units 1 and 2, respectively.
The material for the primary loop fittings is cast stainless steel (SA351-CF8M) for Prairie Island Units 1 and 2.
STAFF EVALUATION CRITERIA The staff's criteria for evaluation of compliance with the revised GDC 4 are provided in Chapter 5.0 of NUREG-1061, Volume 3, and are as follow:
e we
,--y..
. (1) The loading conditions should include the static forces and moments (pressure, deadweight and thermal expansion) due to normal operation, and the forces and moments associated with the safe shutdown earthquake (SSE). These forces and moments should be located where the highest stresses, coincident with the poorest material properties, are induced for base materials, weldments, and safe-ends.
(2) For the piping run/ systems under evaluation, all pertinent information which demonstrates that degradation or failure of the piping resulting from stress corrosion cracking, fatigue, or water hammer are not likely, should be provided.
Relevant operating history should be cited, which includes system operational procedures; system or component modification; water chemistry parameters, limits and controls; resistance of material to various forms of stress corrosion; and performance under cyclic loadings.
(3) A through-wall crack should be postulated at the highest stressed locations determined from (1) above. The size of the crack should be large enough so that the leakage is assured of detection with at least a factor of ten using the minimum installed leak detection capability when the pipe is subjected to normal operational loads.
(4) It should be demonstrated that the postulated leakage crack is stable under normal plus SSE loads for long periods of time; that is, crack growth, if any, is minimal during an earthquake. The margin, in terms of applied loads, should be at least 1.4 and should be determined by a crack stability analysis, i.e., that the leakage-size crack will not experience unstable crack growth even if larger loads (larger than design loads) are applied. This analysis should demonstrate that crack growth is stable and the final crack size is limited, such that a double-ended pipe break will not occur.
. (5) The crack size should be determined by comparing the leakage-size crack to the critical-size crack. Under normal plus SSE loads, it should be demonstrated that there is a margin of at least 2 between the leakage-size crack and the critical-size crack to account for the uncertainties inherent in the analyses, and leakage detection capability.
A limit-load analysis may suffice for this purpose; however, an elastic-plastic fracture mechanics (tearing instability) analysis is preferable.
(6) The materials data provided should include types of materials and materials specifications used for base metal, weldments and safe-ends, the materials properties including the J-R curve used in the analyses, and long-term effects such as thermal aging and other limitations to valid data (e.g., J maximum, maximum crack growth).
The margins cited in the staff criteria are guidelines. Their applicability is dependent upon the conservatism of the analyses performed.
STAFF EVALUATION AND CONCLUSIONS Based on its evaluation of the analyses contained in the licensee's submittals, the staff finds that the licensee has presented an acceptable technical justification, addressing the preceding criteria, to eliminate as a design basis, the dynamic effects of large ruptures in the main loop primary coolant piping of Prairie Island Units 1 an.d 2.
Specifically:
(1) For Prairie Island Unit 1, the loads associated with the highest stressed location in the main loop primary system piping are 2,235 kips (axial),
28,422 in-kips (bending moment) and result in maximum stresses of about 60% of the Service Level D limits specified in Section III of the ASME Code.
For Prairie Island Unit 2, the loads associated with the highest
. stressed location in the main loop primary system piping are 1,623 kips (axial), 28,422 in-kips (bending moment) and result in maximum stresses
)
of about 50% of the Service Level D limits specified in Section III of the ASME Code.
1 1
(2) For the Westinghouse facilities, there is no history of cracking in reactor primary coolant system main loop piping.
The reactor coolant system primary loop has an operating history which demonstrates its inherent stability.
This includes a low susceptibility to cracking from the effects of corrosion (e.g., intergranular stress corrosion cracking),
water hammer, or fatigue (low and high cycle). This operating history totals over 400 reactor-years, including five plants each having 15 years of operation and 15 other plants each with over 10 years of operation.
(3) The leak rate calculations performed for Prairie Island Units 1 and 2 i
used initial through-wall flaws of 7.5 inches and 7.0 inches, respectively and are within the guidelines of NUREG-1061, Volume 3.
Prairie Island Units 1 and 2 have RCS pressure boundary leak detection systems which are consistent with the guidelines of Regulatory Guide 1.45 l
such that leakage of one gpm in one hour can be detected. The calculated I
leak rates through the postulated flaws are large relative to the staff's required sensitivity of the plant leak detection system; the margin is at l
least a factor of ten on leakage for Prairie Island Units 1 and 2.
(4) The margin in terms of load based on fracture mechanics analyses for the leakage-size crack under normal plus SSE loads (Service Level D loads) meets the intent of NOREG-1061, Volume 3, guidance on margins.
Based on a limit-load analysis, the load margin is at least 3 for Prairie Island
. Units 1 and 2.
Similarly, based on the J limit, the margins are about 2 and 1.3 for Prairie Island Units 1 and 2, respectively. Although the margin on the J limit for Prairie Island Unit 2 is less than 1.4 as recommended in NUREG-1061, Volume 3, the staff has determined that if a leakage-size crack slightly less tnan 7.0 inch were assumed, the analysis will meet the 1.4 margin on the J limit, as well as other margins. Thus, the results demonstrated that the margin in terms of load is within the guidelines of NUREG-1061, Volume 3.
(5) The margin between the leakage-size crack and the critical-size crack was calculated by a limit load analysis. The results demonstrated that a margin of about 5 exists for Prairie Island Units 1 and 2 and is within the guidelines of NUREG-1061, Volume 3.
(6) Prairie Island Units 1 and 2 have cast stainless steel piping (and/or fittings) and associated welds in the primary coolant systems. The thermal aging properties of the Prairie Island Units 1 and 2 cast stainless steel materials are described in WCAP-10456 and WCAP-10931 (Revision 1), respectively. As an integral part of its review, the staff's evaluations of the material properties data of WCAP-10456 and WCAP-10931 (Revision 1) are enclosed as Appendices I and II, respectively, to this safety evaluation report. The applied J for Prairie Island Unit 1 in WCAP-10639 for cast stainless steel fittings and 2
associated welds was less than 3,000 in-lb/in and hence the staff's upper bound on the applied J (refer to Appendix I) was not exceeded. The applied J for Prairie Island Unit 2 in WCAP-10929 (Revision 1) for cast stainless steel piping, fittings, and associated welds was less than the maximum allowable J estimated for the specific location and hence the staff's upper bound on the applied J (refer to Appendix II) was not exceeded.
. In view of the analytical results presented in the licensee's submittals, the staff concludes that the probability or likelihood of large pipe breaks occurring in the primary coolant system loops of Prairie Island Units 1 and 2 is sufficier.tly low such that dynamic effects associated with postulated pipe breaks in these facilities need not be a design basis.
Furthermore, the staff concludes that the licensee is in compliance with GDC 4, as revised.
Principal Contributors:
S. Elliot S. Lee Date:
DEC 2 21986 4
d-e l
i l
!