ML20196C616
| ML20196C616 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 12/05/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20196C608 | List: |
| References | |
| NUDOCS 8812080043 | |
| Download: ML20196C616 (8) | |
Text
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION s
j WA3HINGTON, D. C. 20655
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS.107 AND 93 TO FACILITY OPERATING LICENSE NO. NPF-4 AND NFF-7 VIRGINIA ELECTRIC AND POWER COMPANY CLD DOMINION ELECTRIC COOPERATIVE NORTH ANNA POWER STATION, UNITS NO. 1 AND N0. 2 DOCKET NOS. 50-338 AND 50-339
1.0 INTRODUCTION
By letter dated November 6, 1986, and as supplemented by letters dated February 24 and March I?, 1987 and March 8 and June 10, 1988, the Virginia Electric and Power Company (the ?icensee) proposed amendments to Facility Operating Licenses NPF-4 and NPF-7 for the North Anna Power Station, Units No. I dud No. 2 (NA-l&2).
The amendments would permit plant operation with the reactor coolant pump and steam generator sup orts redesigned in accordance with the fracture mechanics 'ivak-before-break" LBB) technology as permitted by the revised General Design Criterion 4 (GDC-4 of Appendix A to 10 CFR Part 50. The amendments would aco a license condition to Operating Licenses NPF-4 and NPF-7 stating that the design of the reactor coolant pump and steam generator supportt m6y be revised in accordance with the licensee's submittal dated November 6,1986 (Serial No. 86-477A).
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 pb11shed in the Federal Register (52 FR 41288) to be effective November 27, 1987, amending G0C-4 of Appendix A to 10 CFR Part 50.
The revised GDC-4 allcws 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 impingenent barriers as well as other related changes in operating plants, plants under construction, and future plant designr. Containment design and emergency core cooling requirements are not influenced by this omdification.
The acceptable technical procedures and criteria are defined in NUREG-1061, Volun.e 3.
Based on the revised GDC-4, the licensee has requested approval for a redesien of the reactor coolant pump and steani generator supports at NA-182.
The revised GDC-4 eliminates the need for consideration of postulated breaks in the reactor coolant system (RCS) prirary loop piping and its effects such as pipe whip, jet impingement, asynnetric pressure loading, and primary component GG12080043 881205 PDR ADOCK 05000338 P
sub-compartment pressurization. Approval of the licensee's request will allow the elimination of certain snubbers which are now required solely to mitigate a pipe rupture event and to replace certain snubbers with rigid restraints which have minim 6', thermal movement.
The licensee's request is based upon the use of advanced fracture mechanics technology as applied to primary system piping in two Westinghouse Topical Reports: WACAP-9558, Revision 2 (May 1981), "Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Through-Wall Crack"; WACAP-9787 (May 1981), "Tensile and Toughness Propert:es of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation"; and Letter Report NS-EPR-2519, E. P. Rhae (Westiaghouse) to 0. G. Eisenhut (NRC) dated November 10, 1981. Approval by the NRC of the above Topical Reports is provided in Generic Letter 84-04 dated February 1, 1984, entitled "Safety Evaluation of Westinghouse Topical Reports Dealing with ~11mination of Postulated Pipe Breaks in PWR Primary Main Loops."
Generic Letter 88-04 provided th NRC staff Safety Evaluation Report for analysis of r.aterials submitted for a group of utilities operating PWR's to resolve Unresolved Safety Issue A-2.
The staff evaluation concluded that provided certain conditions were met, an acceptable technic 1 basis exists 50 that asynetric blowdown loads resulting from large breaks Li main coolant loop piping need not be considered as a design basis. NA-182 were not included with the group of plants for which the Unresolved Safety Issue A-2 Was addressed.
Therefore, to supplement the fracture mechanics studies performed for the A-2 Owner's Group, a plant-specific fracture mechanics study was undertaken for NA-182. Westinghouse reports WCAP-11163 and -11164, dated
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August 1986 and entitled "Technical Bases for Eliminating Large Primary Loop 1
Pipe Rupture as a Structural Design 84. sis for.'lurth Anna, Units 1 and 2," were submitted by the licensee as part of the amendment request dated November 6, 1986.
By letter dated March 8,1988, the licensee submitted WCAP-11163, Supplement 1, entitled "Additional Information in Support of the Technical Justification for E11ninating Large Primary Loop Pipe Rupture as the Structural Design Basis for North Anna Units 1 and 2," dated January 1988, in response to the staff's request for additional information.
The bases for WCAP-11163, Supplement 1 are consistent with the guidelines of NUREri-1061, Volume 3.
In additicn, the licensee referenced the Westinghouse reports WCAP-10456, "The Effects of Thermal Aging on the Structural Integrity of Cast Stainless Steel Piping for Westinghouse Nuclear Steam Supply Systems" dated November 1983, and WCAP-10931, "Toughness Criteria for Thermally Aged Cast Stainless Steel,"
Revision 1, dated July 1986. WCAP-10456 and WCAP-10931 have been previously reviewed by the staff.
As noted above, acceptable technical procedures and criteria are defined in NUREG-1061, Volume 3, and the staff has reviewed and evaluated the licensee's submittal for compliance with the revised GDC-4 A discussion of these matters as well as the staff's findings and evaluation is provided below.
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2.0 OISCUSSION 2.1 NA-182 Primary Locp Piping The NA-182 primary loop piping consists of 34-inch, 36-inch, and 33-inch nominal oianeter hot leg, cross-over leg, and cold leg, respectively. The piping material in the primary ) loops is cast stainless steel (SA-351 CF8A pip bg and SA-351 CF8M fittings.
The piping is centrifugally cast and the fittings are statically cast.
The welding processes used were submerged arc (SAW), shielded tretal ar c (SMAW), and gas tungsten arc (GTAW).
The staff's criteria for evaluation of compliance with the revised CDC-4 are discussed in Chapter 5.0 of Reference 7 and are as follows:
(1) The loading conditions should include the static forces and moments (pressure, deadweight, and thermal expansion) due to normal operation, and the forceF and moments associated With the safe shutdown earthquake (SSE). These forces and moments should be loc 6ted where the highest stresses, coincident with the poorest material properties, are induced for base materials, weldtrents, and safe ends.
(2) For the piping run/ systems under evaluation, all pertinent information which demonstrates that degradation or failure of the piping resulting f rom stress corrosion cracking, fatigue, or water hamer 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; and resistance of material to vai'fous forms of stress corrosion and performance under cyclic loadings.
(3)
The,raterials data provided should include types of materials and materials specifications used for base n.etal, weldrents, and safe ends;
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the materials properties including the fracture mechanics pararreter "J-integral" (J) resistance (J-R) curve used in the analyses; and long-term effects such as thermal aging and other limitations to valid 4
data (e.g., J maximum, and maximum crack growth).
(4) A through-wall flaw should be postulated at the highest stressed locattuns determined from criterion (1) above.
The size of the flaw should be large enough so that the leakage is assyred of detection with 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 trargin, in terms of applieo 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 expertence unstable crack growth even if larger loads (larger than design loads) art applied. However, the final rule permits a reduction of thu margin of 1.4 to 1.0 if the individual normal and seismic (pressure.. deadweight, r
thermal experision, SSE, and seismic anchor motion) loads are sumed absulutely.
This analysis sh0uld demonstrate thdt Crack growth is stable and the final flew size is limited, such that a double-ended pipe break will not occur.
4 (6) The flaw size should be determined by comparing the leakage-size flaw to the critical-size flaw.
Under normal plus SSE loads, it should be cenionstrated that there is a rnargin of at least 2 between the leakage-size flaw ar.o the critical-size flow 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 prrferable.
i The staff has evaluated the information presented in WCAP-11163 and WCAP-11163, Supplement 1 for compliance with the revised GDC-4 Furthermore, i
the staff perfortaed independent flaw stability computations using an elastic-plastic fracture trechanics procedure developed by the staff in NUREG/CR-4572, "NRC Leak-Before-Break (LBB, NRC) Analysis Method for Circumferentially Through-Wall Cracked Pipes Under Axial Plus Bending Loads,"
May 1986.
t On the basis of its review, the st6ff finds the NA-182 priniary loop piping in l
cctnpliance with the revised GDC-4 The following paragraphs in this section present the staff's findings.
l (1) Normal operating loads, including pressure, osadweight, ano 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 plus SSE loads.
In the stability analysis, the individual normal load components were suntned algebraically and the seismic loads were then added absolutely.
In the leak rate analysis, the individual normal load components were surrsred algebraically.
Leak-before-break evaluetions were performed for the limitirg location in the pipir g.
(2) For Westinghouse facilities, there is no history of cracking failure in RCS primary loop piping.
The RCS primary loop has an operating history which demoristrates its inherent stability.
This includes a low suscep"bility tc cracking failure from the effects of corrosion (e.g., intergranular stress corrosion cracking), water horrner, or fatigue (low and high cycle).
This operating history totals over 450 reactor-years, including 5 plants each having over 16 years of operation and 15 other plants each with over 11 years of operation.
(3) The material tensile and fracture toughness properties were provided in WCAP-11163andWCAP-11163)Supplerrent1.
Because there are cast stainless steel piping (and fitting and associated welds in the NA-112 primary loop, the thermal aging toughness properties of cast stainless steel traterials were estimated according tu procedures in WCAP-104Ef and WCAP-10931.
The material tensile properties were estimated using generic procedures.
For flaw stability evaluations, the lower-bound stress-strain properties were used.
For leakage rate evaluations, the average stress-strain properties were used.
.. (4) NA-182 have RCS pressure boundary leak detection systems which are consistent with the guidelines of Regulatory Guide 1.45 such that a leakage of one gallon per minute (gpn,) can be detected.
The calculated leak rate through the postulated flaw is large relative to the staff's required sensitivity of the plant's leak detection systems; the nargin is at least a factor of 10 cn leakage ano is consistent with the guidelir.es of fiUREG-1061, Volume 3.
(5)
In the flaw stability analyses, the margin in terms of load for thE leakage-size flaw under normal plus SSE loads exceeds 1.4 and is consistent with the guidelines of NUREG-1061, Volurre 3.
(6) The margin between the leakage-size flaw and the critical size flaw was also evaluated in the flaw stability analyses.
The margin in terms of flaw size exceeds 2 and is consistent with the guidelines of NUREG-1061, Volun.e 3.
Based on the above, the staff concludes that the NA-1&2 primary loop piping coniplies with the revised GDC-4 according to the criteria in NUREG-1061 Volume 3.
2.2 NA-182 Reactor Coolant Purt.p and Steam Generator Supports Redesign The supporting syctem redesign would permit NA-182 to recuce the nurrber of large tcre hydraulic snubbers at the RCP arid SG supporting systems of the reactor coolant loops (RCLs).
The new design will eliminate two snubbers from every RCP supporting system, four snubbers frcm every SG supporting system, and replace two snubbers with rigid struts in every SG suppurting system; thereby recucing the number of snubbers in each RCL in NA-1&2 by eight.
The technical
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basis for this redesign is the use of "leak-before-break" analyses, which were used to justify elimination of the dynamic effects of postulate 6 pipe ruptures from the design of primary piping systems.
The analyses perfortred show that the redesign will be able to withstand all remaining loadPigs, including those Caused by the SSE and the limiting high enurgy line breaks at branch oczzles.
Specifically, the analytical results indicated that stresses in the RCL piping are Within the UFSAR allowables with adequate tr.orgins of safety.
hA-1&2 each have three RCLs in their reactor coulant syste.r.s.
Each RCL has one RCP and one SG.
Identical designs are used to support the RCPs and likewise for the SGs.
The RCP support is a frame structure with two snubbers installed to restrict rrovement parallel to the cold leg.
The SGs are supported laterally at two levels. The upper SG support consists of four snubbers tangentially arranged around a ring girder.
The lower SG support is a rectangular-cubic structure intercunnected with the RCP support by four snubbers to restrict rnoverrent per-pendicular to the hot leg direction.
It is also mounted with two other snubbers to restrict moven.ent parallel to the hot leg.
The propcsed redesign wculd eliminate the two snubbers parallel to the hot leg from the Ivwer SG support, two snubbers in the RCP support, and two snubbers from the SG-RCP support interconnection.
Further, it will replace the two snubbers in the hot leg direction by twu rigid struts in the upper RCP support.
6-r Westinghouse Reports WCAP-11163 and 11164 provide the basis for the redesign with reduced loading level, which in turn, requires less support rigidity I
during the remaining dynamic events required for the pl6nt's design.
Loadin cutisidered in the redesign are those cause by deadweight, internal pressure,gs thermal movement, seismic events which include the Operating Bases Earthquake c
(OBE) and SSE, and postulated pipe ruptures at nozzles.
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Two independent analyses were performed to verify the adequacy of the i
recesign.
One was performed by Westinghouse using the WESTDYN computer coce 1
to obtain RCL equiptr>ent and piping stresses.
The other one was performed by the Stone & Webster Engineering Corporation (SWEC) using the STARDYN computer 1
code to obtaire support loads.
Both the WESTDYN and the STARDYN computer codes were approved by NRC in 1974.
Both rathematical models used mass and stiffness representations to sitnulate one complete RCL. Three dimensional seismic analyses were perforrreo by using peak bruddening amplified response spectra with equipment domping values of half and one percent for OBE and SSE respectively.
The three directional seismic responses were combined by the square-rout-sum-of-squares (SRSS) roethod. Combination of closely spaced modes was conducted according to the 10% n:ethod recenciended by Regulatory Guide j
1.92 Rev. 1.
Good agreerrent was fcund between the Westinghouse and the SWEC l
results.
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The redundancy of equipment or ccrrponents in a system should be taken into i
4CC0unt when the reliability of equipment or components is considered.
If a failure occurs in a redundant structure, the consequences may not be as seriuus as in a structure with less inherent redundancy.
In the case of the RCL with many snubbers recoved, the remaining snubbers will need to be more reliable since the level of redundancy nas been reduced.
The Itcensee plans to replace the reraining snubber units with hydraulic snubbers manufactured by Taylor Devices.
The service record of thest snubbers has shown no service-criented failures were eser discovered.
Since NA 182 has ccr:nttted to the maintenance practices recorrrrended by the manufacturers, high reliability of snubber performance will be assured.
The staff finds the licensee has l
proposed redesigning the NA-182 SC and RCP supporting systems by applyir.9 4
i 6pproved technology and by using qualified equipment.
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1 3.0 EVALUATION l
The staff has reviewed the information submitted by the licensee and has performed independent flaw stability computations.
On the basis of its l
review, the staff concludes that the NA-1&2 primary loop piping ccmplies with the revised GDC-4 according to the crituria in NUREG-1061, Volume 3.
- Thus,
.I the probability or likelihood of large pipe breaks occurring in the primary coolant system loops of NA-1&2 is sufficiently low such that dynamic effects associated with postulated pipe breaks need not be a design basis.
In addition, the NA-182 proposed redesign at the SG and RCP supporting systems uses approved technology and using qualified equiptront.
Therefore, the staff finds the results of the supporting analyses to be acceptable.
Dased on all l
of the above, the staff finds the proposed redesign of the NA-182 RCP and SG r
I supports to be acceptable and in confortnance with GDC-4 of Appendis A,10 CFR Part 50.
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4.0 ENVIRONMENTAL CONSIDERATION
These amendments involve a change in the installation or use of a facility component located within the restricted area as defined in 10 CFP Part 20.
l The staff has determined that the amerdments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously i
published a proposed finding that the amendments involve no significant hazards consideration and there nas been no public comment on such finding.
Accordingly, i
the amendments meet the eligibility criteria for categorical exclusion set forth
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Pursuant to 10 CFR 551.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the l
issuance of the amendments.
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5.0 CONCLUSION
We have concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be j
endangered by operation in the proposed manner, and (2) such dctivities will be conducted in compliance with the Commission's regulations, and the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Date: December 5, 1988 Principal Contributors:
I S. Lee H. K. Shaw L. B. Engle l
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DATED:
December 5, 1988 l
AMENDMENT NO.107 TO FACILITY OPERATING LICENSE NO. NPF-4-NORTH ANNA UNIT 1 AMENDNENT NO.
93 TO FACILITY OPERATING LICENSE NO. NPF-7-NORTH ANNA UNIT 2 EStetst SileE MS
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NRC & Lo' cal'PDAs i
PDII-2 Reading S. Yarga, 14/E/4 G. Lainas, 14/H/3 H. Berkow i
D. Hiller
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L. Engle I
OGC-WF D. Hegan, 3302 MNBB i
E. Jordan, 3302 MNBB l
B. Grimes, 9/A/2 T. Meek
-(8),P1-137 i
Wanda Jones, P2130A E. Butcher ACRS (10), 11/F/23 GPA/PA ARM /LFHB B. Wilson, R-II S. Lee H. Snaw cc: Plant Service list i
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