ML20086K175
| ML20086K175 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/06/1991 |
| From: | Joshua Wilson TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| IEB-88-011, IEB-88-11, TAC-M72166, TAC-M72167, NUDOCS 9112120281 | |
| Download: ML20086K175 (4) | |
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December 6, 1991 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desi Washington, D.C.
20555 Centlemen:
In the Matter of
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Docket Nos. 50-327 Tennessee Valley Authority
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50-328 SEQUOYAll NUCLEAR PLANT (SQN) - NRC BULLETIN 88-ll, PRESSURIZER SURGE LINE THERMAL STRATIFICATION (TAC NOS.pi72166/172167)
References:
1.
TVA letter to NRC dated January 28, 1991, same subject 2.
NRJ letter to TVA dated August 9, 1991. " Audit of Sequoyah Nuclear Plant Bulletin 88-11 [ssues Regarding Furge Line Thermal Stratification (Tac Nos. 72166 and
/2167)"
The purpose of this letter is to respond to NRC's request fcr l
justification for continued operation as stated in Reference 2.
The l-enclosuri containe the required information, as clarified in a conversation with NRC staff on August 19, 1991, to allow SQN to operate until each urit's Cycle 6 refueling outage. Upon completion of the proposed modifications, each unit will meet the American Society of Mechanical Engineers Code Equation 12 stress limits.
Additional justification for continued operation is provided by Westinghoase Commercial Atomic 1 Power (WCAP) 12777 that was transmitted in Referencs 1.
The WCAP allows for 11 hestup-cooldown cycles between May 1990 and modification implementation. To date, each unit has used 4 of the 11 cycles. The proposed modifications will be implemented during each unit's Cycle 6 outage, and the ll-cycle limit is not expacted to be reached before making the necessary mccifications. As previ6usly committed in Reference 1, should either of SQN's units reach nine heatup-cooldown cycles before performing the necessary modifications, NRC will be informed.
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2 U.S. Nuclear Regulatory Commission
_ December 6, 1991
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-TVA has performed a preliminary' evaluation of the pipe rupture _ protective devices for the pipe rupture loads and displacements prepared by Westinghouse Electric Corporation.- A walk down of the Unit I surge line has'been performed during the current Unit 1 C/cle 5 refueling outage to measure.cctual gaps and identify potential new jet impingement targets.
'j No new targets were identified. A similar walkdown is planned for_the-upcoming: Unit 2 refueling outage. After performing a constructibility evaluation of potential-whip restraint modifications, the design change-package will be revised. The preliminary evaluations indicate that minor J
modifications _will be reqaired to two of-the whip restraints.
Based on the_ preliminary evaluation, TVA is withdrawing its request for approval to 'use the leak-before-break methodology on the pressurizer surge line -
at thisjtime.
Previous commitments, as discussed in Enclosure 5 of Reference 1, are still-in-e fect ex 'pt for modifying the implementation time table (i.e.,
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each unit's Cycle-6. refueling-outage).
Please direct questions concerning this issue to W. C Ludwig at (615) 843-7460.
Sincerely,
[f ; <7s_,
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J L. Wilson.__
Enclosure E
cc (Enclosure):
Mr. D. E. LaBarge, Project Manager U.S. Nuclear Regulatory Conmission one White Flint, North 1555 Rockville Pike
- ckville, Maryland 20852 t
NRC Resident Inspector Sequoyah Nuclear Plant-2600 Igou Ferry Road
- Soddy Daisy, Tennessee 37379 Mr. B.
A'. Wilson, Project Chief U.S.-Nuclear Regulatory Commission Region II 101 Marietta. Street,:NW, Suite 2900
- Atlanta, Georgia 30323 a
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Regarding plant operation, the start-up and shut-down procedures, General Operating Instruction (GOI)-1 (Revision 88), " Plant Startup From Cold Shutdown To Hot Standby," and G01-3 (Revision 49), " Plant Shutdown From Minimum Load To Cold Shutdown," have already been revised to ensure that the operators =are aware of potential surge line stratification and to minim'ze the system Delta T.
In summary, modifications to the whip restraint gaps that will eccommodate the ti irmal disolacement from stratification are currently scheduled for the Cycle 6 refueling outages.
Past experience has shown no evidence of gross structural deformation.
In addition, Equation 12 exceef.ances ha N been included in the overall fatigue usage factor calculation of 0.3.
Therefore, the calculated Equatien 12 exceedance for one additional plant cycle will not result-in a meanlugful effect on surge line structural integrity, m
Referene)s:
1.
A.R.C. Markl, " Fatigue Tests of Piping Components," Trans.
ASME 74, 137-303 (1952) s 2.
WCAP-12639 " Westinghouse owners Group Pressurizer Surge Line Thermal Stratification Generic Detailed Analysis Program MUHP-1091 Summary Report," June 1990
-3, WCAP-12777, " Structural Evaluation of Sequoyah and Wat ' Bar Units 1 and 2 Pressurizer Surge Lines, Considering the Effects of T'.)ermal Stratification," December 1990 l
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ENCLOSURE i
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Sequoyah Nuclear Plant (SQN) - Pressurizer Surge Line Thermal Stratification Bulletin 88-11, Equation 12 Response Thernal expansion, s tratification, and striping will resul t in the full range of stresses that are termed secondary and peak stressen, Limits are impased on accondary stresses to prevent excessive displacements and gross plasticity and on peak stresses to prevent fatigue in the pipe.
The basic characteristics of tnese types of stresses are that they are self-limiting; thus a failure from a single application of a secondary or peak icading is not expected.
For surge line stratificacion, the potential effects of excessive displacement have been investigated through tne detailed visual inspection required by Action 1.a of NRC Bulletin 88-11.
Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code imposes I! nits on secondary stresses resulting from thermal exparsion such that shakedown to elastic action is assured. An evaluation of the SQN pressuricer surge line piping indicated that, by conservative transient assumptions and structural analysis, the code Equation 12 allowable of 35m (i.e., the shakedown limit) was exceeded by 11 percent, However, the walkdown and nondestructive examination af the curge line, based on NRC Bulletin 88-11 requirements, indicated no visible evidence of gross distortion at any locations on the surge line.
Therefore, it is concluded that large plastic deformations have not occurred during past plant operation.
This indicates that either the actual operating tranoients are less severe than the design transients, or conservatism exists in the current code methodology for surge line material such as austenitic stainless steel, or the analytical method is very conservative in predicting piping stress.
Based on Mark 1's ecst data in 1952 (Reference 1), test stresses were higher than the calculated code Equation 12 allowable stresses ter many test specimens, yet, tests resulted in a great deal more cycles thaa the code calculated cycles.
Typically, fatigue failure did not occur until 500 to 800 cycles of stress at the shakedovn limit were experienced.
This fact indicates that the Equation 12 shakedown limit is very consarvative, Equction 12 stress requirementa are also somewhat redundant as the usage factor evaluation is not cased on Eccation 12 stress, but rather en Equation 14 stress.
Furthermore, the amount af plasticity in the pipe was addressed by a penal ty f actor, Kc, in the ASME code evaluation.
In the case of SQN, the calculated fatigue usage factor (0.3) was based on the exis;ing configuration for 11 additional heatup and cocidown cycles before modifications to the restraint are implemented.
Therefore, from a secondary stress and fatigue viewpoint, the plant can be safely operated in the existing structural configuration for tne 11 heatup specified in Westinghouse Commercial Atomic Power (WCAP) 12777 (Reference 3).
The Equation 12 exceedance is also a result of conservative assuaptions made in the operatit.g transients.
First, surge line Delta T is assumed to be 95 percent of the system Delta T, compared with mean values of 70 to 30 per:ent from Westinghouse Owners Group monitoring data (Reference 2).
System Delta T is.the dif ference between pressurizer liquid and Loop 2 hot leg temperatures. Second, the conservative 320 degrees Fahrenhelt' system Delta T used in the analysis nas been reached only six times on Unit 2 and four times on Unit 1-in the pa:t ten years of operation. Third, the asswnption of a step change in temperature at the hot to cold interface at a given diametral cross-section of the surge line is also mry conservativ2.