ML20247B489

Safety Evaluation Supporting Util 890330 Request to Eliminate Dynamic Effects of Postulated Primary Loop Pipe Ruptures from Design Basis of Plant,Using leak-before- Break Technology as Permitted by Revised GDC 4

ML20247B489
Person / Time
Site:	Sequoyah
Issue date:	07/19/1989
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20247B454	List: ML20247B450 ML20247B489
References
NUDOCS 8907240133
Download: ML20247B489 (5)

Text

, _ pug . h

% .og ' UNITED STATES l

y. 3 NUCLEAR REGULATORY COMMISSION

, , y j WASHINGTON, D. C. 20555

( .....

/ ,

i ENCLOSURE j SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ELIMINATION OF POSTULATED. PRIMARY LOOP PIPE RUPTURES AS A DESIGN BASIS TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328  !

l l.0 INTRODUCTION By letter dated March 30, 1989, the Tennessee Valley Authority (the licensee) requested the elimination of the dynamic effects of postulated primary loop pipe ruptures from the design basis of Sequoyah Nuclear Plant, Units 1 and 2, using " leak-before-break" (LBB) technology as permitted by the revised General Design Criterion 4 (GDC-4) of Appendix A to 10 CFR Part 50.

The licensee submitted the technical basis for the elimination of primary loop pipe rupture: for Sequo ah Nuclear Plant Units 1 and 2 in Westinghouse report WCAP-12011 Reference 1. The licensee also referenced Westinghouse reports WCAP-10456 Reference 2 and WCAP-10931, Revision 1 (Reference 3), which have been reviewed previously by the staff as discussed in References 4 and 5 respectively.

The revised GDC-4 is based on the development of advanced fracture mechanics technology using the LBB concept. On October 27, 1987, a final rule was published (52 FR 41288), effective November 27, 1987, amending GDC-4 of Appen-dix'A to 10 CFR Part 50. The revised GDC-4 allows the use of analyses to eliminate from the design basis the dynamic effects of postulated pipe ruptures in high energy piping in nuclear power units. The new technology reflects an engineering advance which allows simultaneously an increase in safety, reduced worker radiation exposures, and lower construction and maintenance costs.

Implementation permits the removal of pipe whip restraints and jet impingement barriers as well as other related changes in operating plants, plants under construction, and future plant designs. Although functional and performance requirements for containments, emergency core cooling systems, and environ-mental gaalification of equipment remain unchanged, local dynamic effects uniquely LBB may~be associated withthe excluded from postulated design basisruptures in piping)which (53 FR 11311 qualified for

. The acceptable technical procedures and criteria are defined in NUREG-1061, Volume 3 (Reference 6).

Using the criteria in Reference 6, the staff has reviewed and evaluated the l licensee's submittal for compliance with the revised GDC-4. The staff's findings are provided below. l

• 8907240133 890719 PDR P

ADOCK 05000327 PDC L - - - - -

2.0 EVALUATION 2.1 Sequoyah primary Loop Piping Sequoyah primary loop piping consists of 34-inch, 37-inch, and 33-inch nominal diameter hot leg, cross-over leg, and cold leg, respectively. The piping material in the primary loops is austenitic cast stainless steel (SA-351 CF8M).

The piping its centrifugally cast and the fittings are statically cast.

2.2 Staff Evaluation Criteria >

The staff's criteria for evaluation of compliance with the revised GDC-4 are discussed in Chapter 5.0 of Reference 6 and are as follows:

(1) The loading conditions should include the static forces and moments (pressure, deadweight, and thermal expansion) due to nomal operation, and the forces and moments associated with the safe shutdown earthquake (SSE). These forces and moments should be located where the highest 4 stresses, coincident with the poorest material properties, are induced '

for base materials, weldments, and safe ends.

(2) For the piping run/ systems upGr evaluation, all pertinent information which demonstrates that dear,idation 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 occrational procedures; system or component modification; water chemistry parameters, limits, and controls; and resistance of material to various forms of stress corrosion and performance under cyclic loadings.

(3) 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 fracture mechanics parameter "J-integral" (J) resistance (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, and maximum crack growth).

(4) A through-wall flaw should be postulated at the highest stressed  !

locations determined from criterion (1) above. The size of the flaw should be large enough so that the leakage is assured of detection with 1 at least a factor of 10 using the minimum installed leak detection capability when the pipe is subjected to normal operational loads.

(5) It should be demonstrated that the postulated leakage flaw 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 flaw stability analysis, i.e., that the leakage-size flaw will not experience l A

1

.#' e' 3

Lunstable crack growth even if larger loads (larger than design loads) are p applied. However, the final rule permits a reduction of the margin of.

L 1.4 to 1.0 if the individual normal and seismic-(pressure,: deadweight,

i. thermal expansion, SSE,.and seismic anchor motion) loads are summed L

absolutely. This analysis should demonstrate that crack ~ growth is stable

and the' final flaw size is limited,. such that a ~ double-ended _ pipe break will not occur.

(6)Theflawsizeshouldbedeterminedbycomparingtheleakage-sizeflawto the critical-size flaw. Under normal plus SSE loads, it should be demonstrated that there is a margin of at least 2 between the leakage-size flaw and the' critical-size flaw 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.

2.3 Staff Evaluation of GDC-4 Compliance The ' staff has evaluated the information presented in Reference 1 for compliance with the revised GDC-4. Furthermore, the staff performed independent flaw stability computations using an elastic-plastic fracture mechanics procedure.

developed by the staff (Reference 7).

On the basis of its review, the staff finds the Sequoyah primary loop piping in, compliance with the revised GDC-4. The following paragraphs in this section.

present the staff's evaluation.

(1) Normal operating loads, including pressure, deadweight, and thermal expansion, were used to determine leak rate and leakage-size flaws. The flaw stability analyses performed to assess margins against pipe rupture at postulated faulted load conditions were based on normal pins SSE loads.- In the stability analysis, the individual normal and seismic loads were summed absolutely. In the leak rate analysis, the individual normal load components were summed algebraically. Leak-before-break evaluations were performed for the limiting location in the piping.

(2) For Westinghouse facilities, there is no history of cracking failure in reactor coolant system (RCS) primary loop piping. The RCS primary loop has an operating history which demonstrates its inherent stability. This includes a low susceptibility to cracking failure from the effects of corrosion (e.g., intergranular stress corrosion cracking), water hammer, orfatigue(lowandhighcycle). This operating history totals over 450 reactor-years, including 5 plants each having over 17 years of operation and 15 other plants each with over 12 years of operation.

(3) The material tensile and fracture toughness properties were provided in Reference 1. Because the Sequoyah Nuclear Plant Units 1 and 2 primary loop piping consists of cast stainless steel, the thermal acing toughness

.f 4

L f' properties of cast . stainless steel materials were estimated according to

i. procedures in References 2 and 3. The material tensile properties were L estimated using plant specific material certifications and generic procedures. For flaw stability evaluations, the lower-bound stress-strain properties were used. For leakage rate evaluations, the averaga stress-strain properties were used.

(4).Sequoyah Nuclear Plant Units I and 2 have RCS pressure boundary leak detection systems which ats :onsistent with the pidelines 'of Regulatory Guide 1.45 such that a leakage of one gallon per .n!nute (gpm) in one hour l can be detected. The calculated leak rate through the postulated flaw is l large relative to the staff's required sensitivity of the plant's leak l ' detection systems; the margin is a factor of-10 on leakage and is l consistent with the guidelines of Reference 6.

(5) In the flaw stability analyses, the staff evaluated the margin in terms of load for the leakage-size flaw under normal plus SSE loads. The staff's calculations indicated the margin exceeded 1.0 when the individual normal and seismic loads were summed absolutely. The margin is consistent with the guidelines of the final rule.

(6) Similar to item (6) above, the margin between the leakage-size flaw and the critical-size flaw was also evaluated in the flaw stability analyses'.

The staff's calculations indicated the margin in terms of flaw size exceeded 2 for the load combination methoc considered. The margin is consistent with the guidelines of Reference 6.

3.0 CONCLUSION

The staff has reviewed the information submitted by the licensee and has performed independent flaw stability computations. On the basis of its review, the staff concludes that the Sequoyah Nuclear Plant, Units 1 and 2, primary loop piping complies with the revised GDC-4 according to the criteria in NUREG-1061, Volume 3 (Reference 6). Thus, the probability or likelihood of

.large pipe breaks occurring in the primary coolant system loops of Sequoyah

'is sufficiently low such that dynamic effects associated with postulated pipe breaks need not be a design basis.

4.0 REFERENCES

(1) Westinghouse Report WCAP-12011. " Technical Justification for Eliminating large Primary Loop Pipe Rupture as the Structural Design Basis for Sequoyah Units 1 & 2," October 1988, Westinghouse Proprietary Class 2.

(2) Westinghouse Report WCAP-10456, "The Effects of Thermal Aging on the Structural Integrity of Cast Stainless Steel Piping for Westinghouse Nuclear Steam Supply Systems," November 1983. Westinghouse Proprietary Class 2.

! (3) Westinghouse Report WCAP-10931 Revision 1, " Toughness Criteria for Thermally Aged Cast Stainless Steel," July 1986, Westinghouse Proprietary Class 2.

s e

'4

• .. e

.e.

w. .

(4) Letter from B.L J. Youngblood of NRC to M.. D. Soence of Texas Utilities Generating Company dated August 28, 1984.

(5) Letter from D.- C. D11 anni of NRC to D.' M. Musolf of Northern States Power Company dated December 22 1986.

(6) NUREG-1061, Volume 3, " Report of the U. S. Nuclear Regulatory Commission Piping 2eview Committee, Evaluation of Potential for Pipe Breaks,"

November 1984.

(7) NUREG/CR-4572, "NRC Leak-Before-Break (LBB.NRC) Analysis Method for

'Circumferentially Through-Wall Cracked Pipes Under-Axial Plus Bending Loads,";.May.1986.

Principal Contributor: S. Lee Dated: . July 19, 1989 l

O

__AL._=._-.._.. __a.--- - - - -