ML20217H268

Application for Amend to License DPR-35,modifying Tech Specs 3.6.A.1 & 4.6.A.1 as It Pertains to Primary Sys Boundary, Thermal & Pressurization Limitations & Surveillance Requirements & Basis 3/4.6.A.Calculation Encl

ML20217H268
Person / Time
Site:	Pilgrim
Issue date:	03/25/1998
From:	Olivier L BOSTON EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20217H272	List: ML20217H268 ML20217H279 ML20217H329
References
LTR.2.98.0023, NUDOCS 9804030248
Download: ML20217H268 (10)

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10CRF50.90 10CRF50.91 g

Boston Edison Pilgrim Nuclear Power Station Rocky Hill Road Plymouth, Massachusetts 02360 LJ. Olivier Vice President Nuclear and Station Director March 25, 1998 BECo Ltr. 2.98.0023 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Docket No. 50-293 License No. DPR-35 Proposed: License Amendment to BECo Technical Specification 3.6.A.1 to Eliminate the 145 F Differential Temperature Limit and Modify 4.6.A.1 Surveillance Requirements and Basis 3/4.6.A.

Boston Edison Company (BECo) hereby proposes to amend Operating License No. DPR-35 in accordance with 10CFR50.90. This proposed change modifies Pilgrim Nuclear Power Station (PNPS) Technical Specification 3.6.A.1 and 4.6.A.1 as it pertains to Primary System Boundary, Thermal and Pressurization Limitations and Surveillance Requirements, and Basis 3/4.6.A.

NRC review and approval of the proposed amendment is requested.

The requested change is described in Attachment A. The marked-up Technical Specification pages are provided in Attachment B. Attachment C contains the amended Technical Specification pages.

L4 Lvv7 . Olivier 9004030248 900325 \l PDR ADOCK 05000293 s

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Commonwealth of M:sscchus;tts)

Country of Plymouth )

Then personally appeared before me, L.J. Olivier, who being duly swom, did state that he is Vice President Nuclear, Station Director of Boston Edison Company and that he is duly authorized to execute and file the submittal contained herein in the name and on behalf of Boston Edison Company and that the statements are true to the best of his knowledge and belief.

My commission expires M w [u! L >c.2<>.JL OCA

/ DATE NOTARY PUBLIC ~

PETER M.KAHLER NOTARY PUBuc Wh EWis Sept 20,2002 Attachments: A. Description of Proposed Change B. Marked-up Technical Specification Pages 3/4.6-1 and B3/4.6-1 C. Amended Technical Specification Pages 3/4.6-1 and B 3/4.6-1 D. Calc. No. M-778, Rev. O, " Evaluation of RPV Flange to Adjacent Shell 145 Deg. Differential Temperature Limit" Regional Administrator, Region 1 Peter LaPorte, Director U.S. Nuclear Regulatory Commission Massachusetts Emergency Management Agency Office of Emergency Preparedness 475 Allendale Road 400 Worcester Road King of Prussia, PA 19406 P.O. Box 1496 Framingham, Ma. 01701-0317 Senior Resident Inspector Pilgrim Nuclear Power Station Mr. Alan B. Wang Project Manager Project Directorate 1-3 Office of Nuclear Reactor Regulation Mail Stop: OWFN 1482 1 White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. Robert M. Hallisey. Director Radiation Control Program Center for Communicable Diseases Mass. Dept. of Public Health 305 South Street Jamaica Plain, MA~ 02130 C_ _ _-- __ . - . . . . . .

ATTACHMENT A Description of Proposed Chanae Elimination of Technical Specification 3.6.A.1 Requirement RPV Flange to Adjacent Shell Differential Temperature PROPOSED CHANGE:

PNPS Technical Specification (T.S.) Section 3.6.A.1, Ref.[2), requires the maximum temperature difference between the (lower) flange and adjacent shell shall not exceed 145 F (when averaged over a one hour interval) during heatup and cooldown transients as measured at the thermocouple (T/C) locations. The T. S. also requires reactor vessel (lower) flange and adjacent shell temperatures be monitored and recorded every 15 minutes during plant heatup and cooldown conditions. An analysis, Ref.(10), was performed to evaluate whether there remains a need for vessel flange to adjacent shell differential temperature (DT) measurements if a peak DT can be calculated which, while satisfying the requirements of the original construction code, preclude the need for a limitation on the DT gradient currently required by the T. S., Ref.[2].

The Ref. [10] analysis demonstrates the DT gradient between the vessel lower flange and adjacent shell is dependent upon vessei heatup and cooldown rate. As long as heatup and cooldown events are restricted to rates that do not exceed 100 F/hr when averaged over any one hour period, there is no need to measure the differential metal temperature since it is redundant to the fluid ramp limit.

The results of this analysis show this requirement is no longer necessary, and a T. S. change should be instituted which removes the requirement that the reactor vessel flange and adjacent shell DT be monitored during heatup and cooldown events and also removes the 145 F DT limit from the T. S. Section 3.6.A.1.

BACKGROUND The reactor vessel closure region consists of an integral upper head and flange, bolting, and integral lower flange and shell. Reactor vessel closure region metal temperature measurements provide an indirect method to assure the associated thermal stresses related to structuralintegrity and fracture toughness remain within ASME Code acceptable limits. The thermal response of the vessellower flange is slower than the response of the adjacent shell (due to variations in geometry and heat transfer characteristics) when subjected to a fluid transient. This lag in thermal response produces differential thermal growth and stresses between the components which make up the reactor vessel closure region. The transient heat transfer analysis performed by the vessel manufacturer, Combustion Engineering (CE), and the results reported in Ref.[3] conclude the thermal response of the reactor vessel metal will result in stresses well below code allowable limits when the fluid heatup/cooldown rates do not exceed the transients described in Ref's. (5) and (6). Thermocouples attached to the RPV outer flange surface and to the outer surface of the adjacent shell monitor actual metal temperature response to heatup/cooldown transients and provide assurance that the maximum DT specified in T. S. Section 3.6.A.1 is not exceeded. Temperature measurements recorded over several outages and test results of Ref. [9] confirm the analytical results of Ref. [3] are representative of actual conditions during heatup and cooldown.

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An cn lysis, R:f.[10), w s perform:d to evalutts wh:th:r th:rs r:mrins a nr:d for v:ss:1 flinge to cdjacent sh:ll t:mperatura m::sur:m:nts if a prrk DT can be calcul tid which, while satisfying the requirements of the original construction code, precludes the need for a limitation on the temperature gradient between vessel lower flange and adjacent shell currently required by T. S. Ref.[2]. The requirements are R nt the lower flange and adjacent shell calculated stress levels and fracture toughness must remain within the conservative ASME Section 111, Ref.[7], code limits.

DISCUSSION The reactor vessel stresa report, Ref.[3), states the secondary stresses developed from the differential thermal growth of the flange and adjacent shell do not exceed the maximum range of stress intensity specified in the ASME B&PV Code Section lli 1965 Edition, when combined with other appropriate primary mechanical loadings. The maximum differential thermal stress associated with the lower flange and adjacent shell was calculated to be 38 Ksi. This stress represents the " worst case design" heatup/cooldown transient provided in this report and is less than 50% of the maximum allowed by code.

Therefore, there is a significant margin between stresses calculated for the lower flange and adjacent shell using the DT's reported in CENC-1139, Ref.[3), and the maximum DT's necessary to reach the maximum code allowable stress of 80 Ksi. The Ref. [10] calculation establishes a maximum DT that can be developed when associated thermal stresses reach code allowable limits.

This analysis, Ref.[10), also determined the relationship between the flange-to-adjacent shell surface temperatures at T/C locations and calculated metal temperatures based on the thermal response of the fluid transient. Note that metal temperature is redundant to fluid temperature because of transient heat transfer relationships.

The results of Ref.[10] establish the maximum DT permitted between the reactor vessellower flange and adjacent shell when the thermal stresses and fracture toughness associated with this temperature gradient reach the limits prescribed by the original construction code (ASME B&PV Code Section Ill, Ref f?]). Tiic DT's are a result of transient heatup and cooldown conditions at the 100 F/hr, T. S. limit, Ref.[2]. The absolute limits for flange to shell DT's determined in Ref.[10] were compared to the predicted DT's at the " worst case" 100 F/hr heatup or cooldown rates obtained from the reactor vessel analysis of record, CENC-1139, Ref.[3). This comparison was used to establish whether the 145 F flange-to-shell DT limit is an unnecessary, redundant operational restriction.

[lETHOD OF ANALYSIS l

The goveming conditions used in this analysis are the same as those used in the CE stress ,

report, Ref.[3]. The specific technique used to perform the anclysis of the closure region is j described in detail in Ref.[3]. Additional details are provided in Ref.[10). j The bounding analytical heatup transient is from 100"F to 546 F at 100 F/hr, with the critical time at end of heatup. Cooldown is at -100"F/hr from 546 F to 375 F when flooding starts. ,

The temperature of the water and steam drop to 330 F in 10 minutes. The bounding analytical

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cooldown transient then continues at-100 F/hr to 100"F The critical time is at end of cooldown. The flanges become covered at 1.81 hours9.375e-4 days <br />0.0225 hours <br />1.339286e-4 weeks <br />3.08205e-5 months <br />, when the temperature has dropped to I

348 F. Due to lack of similarity with any other condition, the cooldown flooding condition is treated as a separate condition.

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Th3 cqurtions used in ths entlys:s of ths fling:d conn:ctions cnd in ths d:t:rmin tion of bolt preload are taken directly from the CE stress report, Ref.[3). Thermal analyses of the ,

"heatup/cooldown" transients are performed directly in this report. The design (mechanical) I loads rouch as pressure, hydrostatic test, and bolt preload are taken directly from Ref.[3] since f these loadings are unaffected by temperature change in the closure region. The stress resulting from the mechanicalloadings is used to establish the maximum range of stress intensity when combined with stresses resulting from differential thermal growth.

Surface temperatures resulting in skin stresses (or peak stresses) are used in evaluations for fatigue which is not a limiting condition for the maximum temperature gradient. Thus peak I radial thermal stresses are not a consideration in Ref. [10) since only secondary stresses associated with temperature effects are of concern.

Heatup rates of greater than 100 F/hr are assumed to be step temperature changes and are combined into a composite condition called ' rapid heating'. Cooldown rates greater than

-100 F/hr are trected in a similar manner and combined under a composite condition called

' rapid cooldown'. Step changes in temperature do not govern when calculating maximum primary plus secondary stress levels since they are isothermal and for short duration.

Consequently the ' Normal Heatup/Cooldown' conditions remain the most severe for primary plus secondary stress calculations. Stresses resulting from bolt loadings are also required for the calculation of the maximum range of stress intensity. Maximum bolt loads were taken from the 'Biach' stud tensioner manual and are 33% higher than the load required for preload.

The bolt loads are described in Ref's.[3] and (10) as " Design Bolt Tension, Design Bolt Preload, Design Pressure Test, Hydro Bolt Tension, Hydro Bolt Preload and Hydro Pressure Test".

Ref. [10) method follows the techniques used in Ref.[3] with one significant modification: the CE stress report analyzed the thermal conditions prescribed by GE Ref's. [5 & 6] and reported the results without regard to maximizing stress levels. During the heatup transient, CE determined the maximum stress associated with differential thermal growth between the lower flange and adjacent shell occurred at the end of the fluid ramp-up and reported "the

} differential thermal stresses developed between the lower flange and adjacent shell reached 38 Ksi". Since there is no stress reversal, this stress intensity was combined with ambient (0 Ksi stress) and recorded in the CE stress report as the maximum range of stress intensity which is less than the 80 Ksi allowed. The Ref. [10) calculation of maximum thermal stress due to transient conditions is, therefore, unaffected by the " peak range calculation" of Ref. [8]

and can be increased until thermal stresses associated with thermal expansion reach and l

override the goveming " Hydrostatic Bolt-up* condition .

To raconstruct the original analysis and preserve the original design basis, the Ref.[10) calculation required the same esiumptions made in the original report. Most design basis assumptions remained in effect when performing the convergence analysis to extrapolate out i to the peak DT gradient between the flange and shell and are consistent with the requirements of the design specification, Ref.[5), and those contained in ASME Sect.lli. Any additional assumptions er modifications to original assumptions are described in the workbook, Ref. [10).

SUMMARY

OF RESULTS and CONCLUSIONS I

Reactor vessel metal temperatures are dependent on the recirculation fluid heatup or cooldown ramp rate which are measured and controlled. The results of this analysis, Ref. [10),  !

show that as long as the fluid ramp rate is under 100 F/hr when averaged over a one-hour period, the thermal response of the reactor vessel metal at the outer surface of the lower 3

fl:ng] cnd cdjr. cent sh:llis within Ecc:ptibts str:ss limits specifi d by ASME S:ction 111. This cnrlysis dso shows thit th3 fr ctura toughn:ss r:quir m:nts sp;cifi d in ASME S:ction til Appendix G have also been met.

The outer surface DT's between the lower flange and adjacent shell, when measured at the T/C locations are as follows:

Flance to Adiacent Shell DT Calculations Per CENC 1139 Ref.I31 DT at T/C's Heatup 74 F Cooldown -83 F Flance to Adiacent Shell Peak Allowable DT Calculations Per M-778 Ref.f101 DT at T/C's Heatup 149 "F Cooldown -166 F The peak outer surface DT allowed between the lower f!ange and adjacent shell, when measured at the T/C locations, was calculated to be 166 F. This temperature is 83 F higher than the DT reported in Ref.[3] and is 21 F higher than the 145 F DT limit stated in the T. S.

Stresses will reach, but remain within, code acceptable limits at this higher DT temperature.

Ref. [10] also concludes the requirement to measure this DT during heatup and cooldown is redundant to fluid ramp measurernents. j l

Note: The locations of the T/C's do not (necessarily) reflect the thermal response of the flange l to adjacent shell for the worst case condition because the locations of the T/C's are arbitrary.

For example, If the distance between lower flange and adjacent shell T/C's were shortened (i.e., spacing less than the current locations), the DT measurement would be lower, and conversely, if the T/C's were placed farther apart the DT would be greater. Hence, the 145 F DT requirement i , relative to specific locations that have not been clearly identified anu defined in either the CE stress report, Ref.[3), or the GE design specification 21 A1110 AB, Ref.[5].

Based on the results of the Ref.[10] analysis, the following conclusions are reached: '

It is unrealistic to expect the reactor vessel shell or flange will reach metal temperatures whose stress levels exceed the values reported for the transients analyzed in CENC-1139, Ref.[3].

The thermal analysis of the vesselis based on conservative heat transfer coefficients, and the maximum allowable rate of heatup/cooldown was used. Two conditions would 5 ave to occur for the reactor vessel temperatures to increase over those reported in Ref.[3] and approach, l but not reach, the limits provided in the Ref.[10] report.

1. The ramp would have to exceed the T.S. limit on ramp rate. Pilgrim is limited to a 100 F/hr maximum rate per license requirements Ref.[2] and cannot exceed this rate.

2. The maximum fluid temperature would have to exceed the T.S. temperature and pressure limit curves for the vessel wall temperature to increase above the Ref.[3]

values. This is also not possible based on the physicallimitations of the plant.

Although the heat transfer conditions may vary (deteriorate) over time, any temperature rise would not be significant.

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in th3 cvant this3 limits era exc::d:d, en evtluttion must be performid to d:t:rmins tha cffect on structur:Iint2grity of tha r2:ctor visstl cnd compon:nts. ASME Cods S:ction XI Appendix E, Ref. [11), provides a method for evaluating an operating event that causes excursions outside these limits.

Ref. [10] demonstrates the thermal stresses resulting from DT at the reactor vessel lower flange to adjacent shell cannot exceed the values reported in CENC-1139, Ref.[3), which reports the maximum DT between the lower flange and adjacent shell (at the T/C locations) is 83*F for a " worst caso" cooldown event. This statement is based on the known physical relationship between metal thermal response to a fluid input that assumes transient heat transfer between bodies is conservative and valid. Since the reactor fluid temperatures are measured and recorded every 15 minutes, and the ramp rate cannot exceed 100 F when averaged over any one hour period, there is no need to measure the flange /shell metal temperatures because they are redundant to the fluid ramp. Also, continual measurement of flange to adjacent shell DT during heatup or cooldown is unnecessary once initial startup testing had documented that actual metal temperatures are representative of the analytical values. The startup test report, Ref.[9], confirmed the analytical results of Ref.[3] are reasonable.

Fracture toughness calculations (Ref/s (8] and [10]) were also performed that show the pressure / temperature limits on plant subcritical-critical heatup/cooldown and hydrostatic testing are unaffected by removal of the 145 F restriction on RPV flange to lower shell DT.

The results of this analysis do not affect the conclusions reached in the design basis analysis for reactor vessel integrity. This analysis determined the upper bound limits on DT for the flange /shell juncture do not change the conclusions reached in tha reactor vessel stress .

I analysis, Ref.[3), the reactor vessel fatigue analysis, Ref.[4], or the fracture toughness analysis, Ref.[8].

Ref.[10] also considers whether the T. S. limitation of 145*F on the maximum DT (permissible between the vessel lower flange and adjacent shell) is based on stress level or fracture toughness since fracture toughness limits described in 10CFR50, Appendix G, must not be exceeded. Fracture toughness is also dependent on stress and temperature. Brittle fracture, however, is generally a consideration for the closure region only during boltup at ambient temperature. However, a review of the basis for the T. S. reactor vessel pressure-temperature limits (Ref.[8] Sect. 3.6) indicates the closure region may govern for temperature and stress I conditions other than " Cold" boltup. The "Floodup" event described in Ref.[3] was also considered as a basis for the T. S. limit but was not found limiting in either Ref.[3] or Ref. [10].

The rationale for the flange to adjacent shell DT of 145 F was determined to be based on ,

stress levels associated with thermal growth. These stress levels are limited by code which I effectively limits the stress intensity values associated with fracture toughness. j The Ref.[10] analysis also reiterates the criteria of Ref.[3] that any DT gradient between the  !

lower flange and adjacent shellis dependent upon the plant heatup/cooldown rate. Excessive DT's cannot develop between the RPV lower flange and adjacent shell as long as plant heatup/cooldown events are restricted to rates no greater than 100 F/hr when averaged over a one-hour period. Excessive reactor fluid ramps are precluded by station procedures that control and restrict plant heatup/cooldown rates to less than 100 F/hr.

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R::ctor vsssal m:tl t mparatura is a predictabla responso to fluid transinnt conditions based on classical heat transfer; therefore, actual temperature measurements should approximate temperatures reported in the design basis analysis, Ref.[3]. Initial startup testing results reported in Ref.[9] are comparable to the results for the 100 F/hr fluid ramp reported in Ref.[3].

The maximum DT between the lower flange and adjacent shell reported in Ref.[3] is 83 F for the 100*F cooldown condition, and the results from initial startup testing show 90 F for a similar cooldown condition. Since PNPS cannot procedurally exceed the 100 F/hr restriction, the limiting 145 F DT T. S. requirement cannot be reached.

The PNPS normal heatup and cooldown conditions that occur during scheduled or forced outages are roughly 20 to 40 F/hr. Thus, the temperature measurements taken for these outages have not approached the maximum conditions of 100 F/hr. Temperature measurements taken from initial startup testing, Ref.[9], however, provide the most accurate results available for the RPV lower flange to adjacent shell DT due to fluid ramp rates approaching 100"F/hr. This report states," The maximum rate of temperature change obtained was 90"F/hr during the Residual Heat Removal shutdown cooling test of July 7, 1972." The maximum temperature difference between the reactor vessel adjacent to the flange and the flange during startup was 90 F, which occurred during the heatup on July 11,1972." The 90 *F DT is within reasonable agreement with the analytical results reported in Ref.[3].

Removal of this T. S. requirement does not constitute a safety question since the reactor vessel pressure boundary and ability to maintain a floodable core are not jeopardized as a result of this change.

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION

a. The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The recent analysis, Ref.[10], has shown design and licensing bases for reactor vessel integrity will be maintained, and results supporting the T. S. change show the conclusions reached remain unchanged from previous conclusions reached in Ref.[3]

and as described in the FSAR, Ref.[1]. Structuralintegrity for design basis loading conditions is assured, based on the results of Ref.[10]. The ability to control plant heatup and cooldown rates has been shown by analysis to be unaffected by the removal of this T. S. requirement. This has been confirmed by initial startup testing results and the past 25 years of service.

b. The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

T/C's used to monitor reactor vessel flange to adjacent shell DT are used only during normal startup and shutdown conditions, and removal of the T. S. requirement to monitor this differential temperature will have no affect on the design basis accident conditions. Moderator temperature and pressure are monitored and, in the event fluid ramp rates exceed design basis requirements, an evaluation must be performed to determine the effect on structural integrity of the reactor vessel and components.

ASME Code Section XI, Appendix E, Ref. [11], provides a method for evaluating an operating event that causes excursion outside these limits.

c. The proposed amendment does not involve a significant reduction in the margin of )

safety, j i

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Str:ss and fr:ctura toughn:ss calculations, R:f.[10), h:va shown removal of ths T. S. DT requirement will not increase levels above the conservative design basis limits previously established in the analysis of record, Ref.[3), or those stated in the FSAR, Ref.[1]. The loadings used to determine stresses are the same provided by the original equipment designer and manufacturer. The calculated stress levels and fatigue damage assessment for the existing condition are essentially unchanged from the values reported in the reactor vessel analysis of record, Ref.[3). The results of the recent analysis, Ref.[10), show that the margins of safety, as defined in the bases for the Pilgrim T. S. and the FSAR, are not reduced and vessel integrity will be maintained during all normal and transient conditions previously analyzed and reported in the FSAR.

REFERENCES

[1] FSAR Sections 4.2 and 7.8 " Reactor Vessel Mechanical Design and Instrumentation"

[2] PNPS T. S. Section 3.6 A.1 " Primary System Boundary Thermal and Pressurization Limits"

[3] CE Report No. CENC 1139 " Analytical P.eport for Pilgrim Reactor Vessel" Dated 3/9/1971

[4] Altran Technical Report No. 93177-TR-03 Vol's I thru IV " Pilgrim Reactor Vessel Cyclic Load Analysis" Dated August 1994 (EUDDS/RF 94-101)

[5] GE-APED Specification 21A1110 A9 Dated. Aug. 19,1970.

[6] BECo Drawing (formerly GE Drawing ) M1 A12-2 " Reactor Vessel Thermal Cycles"

[7] ASME Boiler and Pressure Vessel Code Section lil 1965 Edition, including January 1966 addenda and vanous Code Cases

[8] Teledyne Technical Report TR-2318 Dated June 2,1976 " Evaluation of Pilgrim Reactor Vessel Boundary to Appendix G, ASME Code Section lil

[9] NEDO 20262 "Startup Test Results- Pilgrim Nuclear Power Station" Dated. Feb.1974

[10] NESG Calculation M-778 " Evaluation of RPV Flange to Adjacent Shell 145 F Differential Temperature Limit " D ated. Feb. 6,1998

[11) ASME Boiler and Pressure Vessel Code Section X1,1989 Edition, including addenda and ,

various Code Cases.

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ATTACHMENT B Marked-Up Technical Specification Paaes 3/4.61 and B 3/4.6-1

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