B13264, Forwards Revised Preliminary Analysis Summary Re Pressurizer Surge Line Thermal Stratification,Per NRC Bulletin 88-011. NDE Insp History Demonstrates Present Day Integrity of Westinghouse PWR Pressurizer Surge Lines

Forwards Revised Preliminary Analysis Summary Re Pressurizer Surge Line Thermal Stratification,Per NRC Bulletin 88-011. NDE Insp History Demonstrates Present Day Integrity of Westinghouse PWR Pressurizer Surge Lines

ML20244B930
Person / Time
Site:	Millstone, Haddam Neck, 05000000
Issue date:	06/02/1989
From:	Mroczka E, Sears C CONNECTICUT YANKEE ATOMIC POWER CO., NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
B13264, IEB-88-011, IEB-88-11, NUDOCS 8906140050
Download: ML20244B930 (10)

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NAST UTILITIES cenerei Orrices . seiden street. Bernn. Connecticut l =wia wn. cam ns iacme co*" P.O. BOX 270 mcs.,e . cia me w** H ARTFORD. CONNECTICUT 06141-0270  ;

k k J N CcNr'.N.7, (203) 665-5000 June 2, 1989 Docket Nos. 50-213 50-423 i B13264  !

Re: NRC Bulletin No. 88-11 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555

References:

(1) NRC Bulletin No. 88-11: Pressurizer Surge Line Thermal Stratification, dated December 20, 1988.

(2) E. J. Hroczka letter to U. S. Nuclear Regulatory Commission, " Pressurizer Surge Line Thermal Stratification," dated April 28, 1989.

(3) J. F. Stolz letter to E. J. Hroczka, " Schedular Relief Request Related to NRC Bulletin 88-11, Pressurizer Surge Line Thermal Stratification" Item 1.b for Haddam Neck and Millstone Unit 3 (TAC Nos. 72136 and 72145),

dated May 19, 1989.

Gentlemen:

Haddan Neck Plant Hillstone Nuclear Power Station, Unit No. 3 NRC Bulletin No. 88-11 Pressurizer Surge Line Thermal Stratification Reference (1) directs Licensees to perform inspections and analyses to determine if pressurizer surge lines are subject to previously unanalyzed stresses from thermal stratification and thermal striping which may reduce the life of the surge line, and to take appropriate corrective measures. Westinghouse, on behalf of the Westinghouse Owners' Group (V0G), is performing the analyses required by Reference (1). The scope of this work includes the Haddam Neck Plant and Hillstone Unit No. 3.

Preliminary results, provided by Westinghouse, were included as Attach-ment I to Reference (2). That submittal represented our response to Item 1.b in Bulletin No. 88-11.

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On April 11, 1989, Westinghouse met with the NRC Staff on this issue. As a result of that meeting, several minor changes to the preliminary analysis summary were made. The purpose of this letter is to transmit the revised preliminary analysis summary. As such, this letter super-sedes Reference (2). Attachment 1 provides the revised summary with changes noted by change bars. No changes to our schedule for completion of Items 1.a, c, and d of Reference (1) required actions are anticipated.

In Reference (3), the NRC Staff indicated that our proposed approach to this issue is not acceptable. Based upon that letter and discussions with the NRC Staff, it appears that some clarifications are in order.

Reference (3) indicates that the proposed schedule for completing Item 1.b of Reference (1) is not acceptable. We believe that, for the Haddam Neck Plant and Millstone Unit No. 3, an adequate response for Item 1.b was provided in Reference (2). Item 1.b requires analysis to determine if the pressurizer surge line meets applicable requirements for licensed plant life when thermal stratification and striping are factored in. If not, a Justification #or Continued Operation (JCO) or cold shutdown is called for. In lis J of that, the V0G Program conducted the initial analysis showing that the applicable requirements were met in the short term and used it as the basis for a "JC0" which is applicable to Haddam Neck and Millstone Unit No. 3. We have transmitted that "JC0" in Reference (2) and revised it here, thereby meeting the intent of Item 1.b. However, we believe the use of the term "JC0" is inappropriate and so have not titled the attachment as such.

As Attachment 1 indicates, Haddam Neck and Millstone Unit No. 3 vill, with a high degree of confidence, meet the applicable codes at least until Items 1.c and 1.d of Reference (1) are completed. No "JC0" is necessary because no equipment is, in any sense of the word, "inoper-able", nor does any defect currently exist at either plant. We believe this interpretation is consistent with Staff guidance communicated during the April 1989 Regulatory Information Conference during which a paper was presented on the subject of " Equipment Operability - Promptly Determining Operability and Establishing Corrective Action Plans for Degraded or Nonconforming Safety Equipment." Since the equipment of interest regarding Bulletin 88-11 is currently fully operable, there is no need to document a JCO, which is normally reserved for equipment not fully oper-able for some reason. Ve also wish to emphasize that the above comments do not conflict with the Staff's intent to have us document why continued operation is acceptable. That is provided in Attachment (1). Our intent is to use the term "JC0" with increased precision.

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.. June 2, 1989 If ther'e are any questions, do not hesitate to contact my staff directly.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY CONNECTICUT YANKEE ATOMIC POWER COMPANY i M . B. M .

t E. J. Mroczka i Senior Vice President b

By: C. F. Sears Vice President STATE OF CONNECTICUT )

) ss. Berlin COUNTY OF HARTFORD )

Then personally appeared before me,. C. F. Sears, who being duly sworn, did state that he is Vice President of Connecticut Yankee Atomic Power Company and Northeast Nuclear Energy Company, Licensees herein, that he is authorized to execute and file the foregoing information in the name and on behalf of the Licensees herein, and that the statements contained in said information are j true and correct to the best of his knowledge and belief.  !

h %. Da dCt41 o Notary Public cc: W. T. Russell, Region I Administrator D. H. Jaf fe, NRC Project Manager, Millstone Unit No. 3 A. B. Vang, NRC Project Manager, Haddam Neck Plant V. J. Raymond, Senior Resident Inspector, Millstone Unit Nos.1, 2, and 3 J. T. Shedlosky, Senior Resident Inspector, Haddam Neck Plant 1

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Docket Nos. 50-213 50-423 B13264 ATTACHMENT 1 Haddam Neck Plant Hillstone Nuclear Power Station, Unit No. 3 NRC Bulletin No. 88-11 Pressurizer Surge Line Thermal Stratification Preliminary Analysis Results l

June 1989 l

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' U. S. Nuclear Regulatory Commission B13264/ Attachment 1/Page 1 l i

June 2, 1989

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Haddam Neck Plant 1 Millstone Nuclear Power Station, Unit No. 3 NRC Bulletin No. 88-11 Pressurizer Surge Line Thermal Stratification Preliminary Analysis Results A. BACKGROUND It was first reported in INPO SER 25-87 that temperature measurements at a German PVR indicated thermal transients different than design. Recent measurements at several domestic PVRs have indicated that the temperature difference between the pressurizer and the hot leg results in stratified flow in the surge line, with the top of the flow stream being hot (pressurizer temperature) and the bottom being colder (hot leg temperature). The top-to-bottom temperature difference can reach 250' to 300*F in certain modes of operation, particularly Modes 3, 4, or 5 during heatup and cooldown.

Surge line stratification causes two effects:

1. Global bending of the pipe is different than that predicted in the original design.

2. Fatigue life of the piping could be reduced due to the global and local stresses from stratification and striping.

More recently, the NRC has issued Bulletin No. 88-11, " Pressurizer Surge Line Thermal Stratification," dated December 20, 1988, identifying the following actions to be taken by licensees:

1. Conduct visual inspection--valkdown.

2. Update stress and fatigue analysis to account for stratification and striping.

3. Obtain monitoring data, as necessary.

The bulletin encourages licensees to perform Actions 2 and 3 above through collective efforts with other plants. In Oc tober 1988, NNECO, CYAPCO, and other members of the Westinghouse Owners' Group (V0G) authorized a program to perform a generic evaluation of surge line stratification in Westinghouse PVRs that vill address portions of Bulletin No. 88-11.

The V0G program is designed to benefit from the experience gained in t'he performance of several plant-specific analyses on Westinghouse PVR pressurizer surge lines. These detailed analyses included definition of revised thermal transients (including stratification) and evaluations of pipe stress, fatigue usage factor, thermal striping, fatigue crack growth, leak-before-break, and support loads. The overall analytical approach used in all of these analyses has been consistent and has been I

reviewed in detail by the NRC staff.

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U. S. Nuclear Regulatory Commission B13264/ Attachment 1/Page 2 June 2, 1989, As of March 1989, plant-specific analyses have been performed on five domestic VestinFhouse PVRs. In addition, 12 Westinghouse plants have completed or are curreatly performing an interim evaluation of surge line stratification which includes finite element structural analysis of their specific configuration under stratified loading conditions.

B.- WOG PROGRAM STATUS As part of the current V0G program, surge line physical and operating data l have been collected and summarized for all domestic Westinghouse PVRs (55 units). Information relating to piping layout, supports and restraints, components, size, material, operating history, etc., has been obtained.

This data has been evaluated in conjunction with available monitoring data and plant-specific analyses performed by Westinghouse. The results 'of this evaluation vere presented to the NRC in a meeting on April 11, 1989.

.The evaluation is being formalized into a Westinghouse topical report (VCAP 12277, Proprietary and VCAP-12278, Non-proprietary version) scheduled for submittal to the NRC on June 15, 1989.

This topical report forms the bases for the following preliminary results.

j C. PRELIMINARY RESULTS

1. Stratification Severity L Thermal stratification (DT) 100*F) has been measured on all surge lines for which monitoring has been performed and which have been reviewed by the V0G to date (eight surge lines).

The amount of stratification measured and its variation with time (cycling) varies. This variation has been conservatively enveloped, and applicability of these enveloping transients has been demonstrated for plant-specific analyses.

Various surge line design parameters vere tabulated for each plant.

From this, four parameters judged to be relatively significant were identified as:

a. Pipe inside diameter l b. Piping slope (average)

c. Entrance angle of hot leg nozzle

d. Presence of midline vertical riser These parameters were used in a grouping evaluation which resulted in the definition of ten monitoring groups corresponding to various combinations of these parameters at Westinghouse PVRs.

Approximately 40 percent of the plants fall into one group for which a large amount of monitoring data Fas already been received and for which the enveloping thermal transients discussed above are

U.' S.- Nuclear' Regulatory Commission Bi3264/ Attachment 1/Page-3 ,

June 2, 1989 applicable. The remaining 60 percent of Westinghouse PVRs are .l divided among the other nine additional groups. Although monitoring data has not been - received representative of all these groups, in

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general, the combination of significant parameters of these nine groups is expected to decrease the severity of stratification below that.of.the enveloping transients. This conclusion is also supported by a comparison of available monitoring data.

2. Structural Effects Significant parameters which'can influence the structural effects of' stratification are:

a. Location and design of rigid supports and pipe whip restraints

b. Pipe layout geometry and size

c. Type and location of piping components Although the material and fabrication techniques for Westinghouse surge lines are reasonably consistent and of high quality, the design parameters listed above vary among Westinghouse PWRs. This l variation in design is primarily a result of plant-rpecific routing requirements.

A preliminary evaluation, comparing the ranges of these . parameters to those of plants for which plant-specific analysis and interim evaluations are available (approximately 20 percent of Westinghouse PVRs), has been performed. This comparison indicates a high degree of confidence that from a combined transient severity and structural effects standpoint, the vorst configuration has most likely been evaluated. This conclusion is supported by plant-specific analyses covering five plants and interim evaluations of six additional plants (interim evaluation is in progress on'six more plants as of March 1989). These analyses and evaluations have included various piping layouts, pipe' sizes, support and. restraint ' designs, and piping components. Although the full range of variation in these parameters has not been evaluated, experience gained from these evaluations indicates that further evaluations vill not result in a more limiting configuration than those already evaluated.

l l 3. Operating Procedures The V0G currently has available the surveys of operating procedures performed in support of existing plant-specific analyses.

Experience indicates that heatup and cooldown procedures have a significant effect on stratification in the surge line. All conclusions reached by the V0G to date have assumed a steam bubble mode heatup and cooldown procedure which may result in a temperature dif ference between the pressurizer and reactor coolant system (RCS)

l-U [U.S.NuclearRegulatoryCommission B13264/ Attachment 1/Page 4 June 2, 1989, hot leg of more than 300'F. In many cases, individual plant operating procedures and technical specifications provide limits on this value. It is also known that some procedures utilize nitrogen, during at least part of the heatup/cooldown cycle, as a means of providing a pressure absorbing space in- the pressurizer. Based on information currently available to the V0G, a high confidence exists-that the steam bubble mode 'heatup assumed to date is conservative

with respect to westinghouse PVRs.

4. Pipe Stress and Remaining Life The design codes for surge line piping have requirements for checking.

pipe stress limits and the effects of fatigue loadings. These stress limits provide a means of controlling stress from primary loads such as pressure, deadweight, and design mechanical loading, as well as stress from' secondary loads such as thermal and anchor. motion effects.

Stratification in the surge line is a secondary load which vill only affect the qualification of secondary stresses. The qualification of ,

primary stresses is not affected by this loading.

Secon/ary stresses are controlled to prevent excessive displacements and gross plasticity and to prevent excessive fatigue loadings in the pipe. The basic characteristic of a secondary stress is that it is self limiting; thus, a failure from a single application of a secondary loading is not expected.

For the stratification issue, the potential effects of. excessive displacements will be investigated at the Haddam Neck Plant _through a detailed visual observation of the surge line during the walkdown required per Bulletin 88-11 action 1.a, as is noted below in Section

6. These effects have been investigated at Millstone Unit No. 3, as is noted below in Section 6.

The effects of secondary stresses on the remaining life of the surge line have been evaluated on a generic basis through the V0G program.

The following summarizes the results of this evaluation.

All plant-specific analyses performed as of March 1989 have demonstrated compliance with applicable ASME Codes and a surge line fatigue life in excess of a 40-year plant life. Review of plant-specific fatigue calculations indicates that the surge line fatigue life is primarily dependent on the number of heatup and cooldown cycles, rather than years of operation.

Considering the worst-case years of operation (28.5 years) in combination with the worst-case number of heatup/cooldown cycles (75 at a different plant) at any Westinghouse PVR, and assuming a 40-year life for all surge lines, it is estimated that no more than approximately 50 percent of the fatigue life has been used at any Westinghouse plant to date.

[ U. S. Nuclear Regulatory Commission B13264/ Attachment 1/Page 5 June 2, 1989 f

For a design life considering 200 heatup/cooldown cycles (used in plant-specific analyses), this vould indicate approximately 100 remaining cycles. This number of remaining cycles far exceeds the postulated worst-case number for the 2-year time frame needed to resolve the stratification issue.

5. Leak Before Break All the plant-specific analyses performed to date that havc included the loadings due to stratification and striping have validated the

" leak-before-break" concept and have substantiated a 40-year plant life. Fatigue crack growth calculations performed as part of these plant-specific analyses have demonstrated that any undiscovered crack as large as 10 percent of the vall thickness vould not grow to cause leakage within a 40-year plant life. Nevertheless, any postu-lated through-vall crack propagation would most likely result in

" leak-before-break" and thus permit a safe and orderly shutdown.

6. Inspection History The NDE inspection history at the Haddam Neck Plant and Millstone Unit No. 3, as well as all other domestic Westinghouse-designed PVRs, has not revealed any service-induced degradation in the surge line piping that has been attributed to thermal stratification.

CYAPCO vill perform a visual inspection (ASME Section XI, VT-3) of the Haddam Neck Plant pressurizer surge line during the next available cold shutdown exceeding 7 days duration. A plar.ned cold shutdown exceeding 7 days duration is scheduled for the fall of ,

1989. CYAPCO vill submit a letter to the staff, within 30 days after completion of this inspection, per Reporting Requirements 1 and 3 of Bulletin No. 88-11. Should the results of this inspection indicate any gross discernible distress or structural damage, this letter vill address the need for corrective actions, repair, plant-specific analysis and/or monitoring, etc., as appropriate.

NNECO performed a visual inspection (ASME Section XI, VT-3) of the Millstone Unit No. 3 surge line in February of 1989. The results of this inspection did not indicate any gross discernible distress or structural damage in the surge line. These results are available for inspection.

D.

SUMMARY

OF CONCLUSIONS FROM V0G PROGRAM Based on information assembled on surge lines for all domestic Westinghouse PVRs, and evaluation of that information in conjunction with plant-specific and other interim evaluation results, the V0G concludes that:

' . - ' 'U. S.- Nuclear Regulatory Commission B13264/ Attachment 1/Page 6 June 2',1989,

,. 1. A high ' degree of confidence exists that further evaluation vill

l. confirm that the worst combination has already been evaluated for l stratification severity, structural effects, and operating 1 procedures.

2. All plant-specific analyses to.date have demonstrated a"40-year life of the surge ' line. Assuming that further evaluation' leads to the same conclusion. for the remaining Westinghouse PVRs, the vorst-case

-remaining life is approximately 100 heatup/cooldown cycles.

3. Through-vall crack propagation is highly.unlikely; however,

" leak-before-break" would permit . a safe and orderly shutdown if a through-vall leak'should develop.

4. NDE inspection history demonstrates the present day integrity of Westinghouse PVR pressurizer surge lines.

5. While additional monitoring, analyses, and surveys of . operating procedures are expected- to further  ::ubs tantia te ' the- above conclusions, the presently available information on surge line stratification indicates that Westinghouse PVRs may be safely operated while additional data is obtained.

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E. OVERALL CONCLUSION Based on the above discussions, CYAPC0 and NNECO believe it is acceptable

for the . Haddam Neck ' Plant and Hillstone Unit No. 3 to continue power operation for at least ten additional heatup/cooldown cycles. CYAPCO and NNECO have committed to address the requirements of Bulletin No. 88-11 by January 1991.