ML17285B103

Safety Evaluation Supporting Amend 77 to License NPF-21

ML17285B103
Person / Time
Site:	Columbia
Issue date:	03/01/1990
From:	Office of Nuclear Reactor Regulation
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ML17285B102	List: ML17285B101 ML17285B103 ML17285B104
References
NUDOCS 9003190125
Download: ML17285B103 (20)

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0.77 TO FACILITY OPERATING LICENSE NO.

NPF-21 WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO. 2 DOCKET NO. 50-397

1.0 INTRODUCTION

By letter (Ref. 1) dated December 15, 1987, Washington Public Power Supply System (the licensee) proposed an amendment to the Technical Specifica-tions to allow operation of Washington Nuclear Project, Unit 2 (WNP-2) with final feedwater temperature reduction (FFTR) and subsequent thermal coastdown to 65% power in order to increase electrical generation by extendinq the operatinq cycle beyond a 12-month fuel cycle.

The changes to the Technical Specifications include:

(1) addition of a section which limits feedwater temperature when temperature is being reduced to extend the fuel'cycle, (2) chanqe to the minimum critical power ratio (MCPR) limits in Table 3.2.3-1, to specify the operatinq limits avoidinq plant operation in the reqion which may result in a fuel failure during transients, and (3) changes to bases and definitions in the Technical Specifications to incorporate updated descriptions.

By letters dated September 14, 1988 (Ref. ?)

and February 14, (Ref. 13) 1990 the request was supplemented with supportinq information requested by the staff.

By letters dated March 7, 1988 (ref. 9) and April 12, 1988 (ref. 10), the licensee submitted an application for the Cycle 4 fuel reload.

That reload application included limits for FFTR at the end of Cycle 4.

By letter dated March 3, 1989, (Ref. 8) April 20, 1989 (Ref. 11) and June 1,

1989 (Ref. 12) the licensee submitted analyses supportinq a request to amend the license for the fifth cycle of operation.

The support material included analyses of safety limits for final feedwater temperature re-duction.

In support of these proposed

changes, the licensee submitted a General Electric (GE) report, NEDC-31107, entitled "Safety Review of WPPSS Nuclear Project No.

2 at Core Flow Conditions Above Rated Flow Throughout Cycle 1

and Final Feedwater Temperature Reduction."

The GE report presents the results of a safety and impact evaluation for the operation of the WNP-2 with FFWTR and increased core flow (ICF).

The limiting normal operational transients described in the Final Safety Analysis Report (FSAR) have been reevaluated.

The loss-of-coolant accident, fuel loadinq error accident, rod drop accident, and rod withdrawal error event were also re-evaluated.

In addition, the effect of increased pressure differences on the reactor internals components, fuel channels and fuel bundles was also analyzed to show that the design limits will not be exceeded.

The effect of ICF on the flow-induced vibration response of the reactor internals was also 1

9003190125 900301 PDR ADOCK 05000397 PDC

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evaluated to ensure that the response is within acceptable limits.

The thermal-hydraulic stability was evaluated for ICF/FFMTR operation, and the increase in the feedwater nozzle and feedwater sparger usage factors due to the feedwater temperature reduction was determined.

The impact of ICF/FFHTR operation on the containment LOCA response was also analyzed.

The staff has reviewed the proposed changes to the Technical Specifica-tions and the supportinq analytical results (Refs.

2 through 4 and 7) and has prepared the followinq evaluation.

2.0 EVALUATION hlNP-2 is currently operatinq in Cycle 5 with a mixed core consistinq of GE and Advanced Nuclear Fuels Corporation (ANF-formerly Exxon Nuclear) fuels.

Since the ANF fuel has operated fewer cycles than the GE fuel, the ANF fuel assemblies have a higher power peakinq factor due to lower burndown effect.

However, the licensee performed safety analyses based on both types of fuels provided by ANF and GE to establish the operating MCPR limits for each type of the fuel to support the chanqes to the Technical Specifications submitted for NRC review and approval.

The objective of the staff review is to confirm that the thermal-hydraulic desiqn of the core has been accomplished using acceptable

methods, and provides an acceptable margin of safety from conditions which could lead to fuel damaqe durinq transient conditions, and is not susceptible to thermal-hydraulic in-stab ility.

The staff has reviewed the followinq areas:

(1) the safety MCPR limit, (2) the operating MCPR limits (3) thermal-hydraulic stability, (4) reactor vessel beltline materials, (5) thermal fatigue stresses to feedwater

nozzles, (6) reactor internals load impact, (7) flow-induced vibration, (8) feedwater nozzle and feedwater sparqer fatigue usaqe, and (9) the proposed chanqes to the Technical Specification.

2.1 Safety MCPR Limit The safety MCPR limit fuel rods in the core durinq transients.

A NRC for the licensinq The safety MCPR limit tion at conditions of acceptable.

has been imposed to assure that 99.9 percent of the are not expected to experience boilinq transition safety MCPR limit of 1.06 was previously approved by calculations to operate the current Cycle 4 fuel.

of 1.06 is also used for analyses to support opera-final feedwater temperature reduction (FFTR) and is 2.2 0 eratinq MCPR Limits Various transients could affect the operatinq MCPR limits during the extended Cycle 4 operation.

The most limitinq events

( load rejection without bypass (LRNB) and feedwater controller failure (FWCF)) have been analyzed by the licensee to determine which event could potentially induce the largest reduction in the initial critical power ratio (RCPR).

The RCPR values oiven in Table 3.3 of References 4 and 7 are cycle specific values for the Cycles 3 and 4 fuels usinq the COTRANSA (Ref.

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The operatinq NCPR values are determined by adding the RCPRs to the safety limit HCPR.

The maximum MCPR values for different types of fuel (resultinq from limitinq event, LRRB) are specified as the operatinq NCPR limits and are incorporated into Table 3.2.3.1 of the Technical Specifications.

The operating MCPR limits as a function of the core flow were also calculated and tabulated in Table 2.1 of Reference 3.

Since the approved method was used to determine the operatinq HCPR limits to avoid violation of the safety limit MCPR in the event of any anticipated transients, we conclude that these limits are acceptable for operation of the extended Cycle 4 fuel.

Based on the review of the licensee's calculations in References 4 and 7

for fuels of the Cycles 3 and 4, we have also determined that the larqest calculated CPR changes due to the effect of FFTR are acceptable for future extended reload applications with operation of FFTR provided that the assumptions made in performinq transient analysis remain consistent with that in Reference 4.

The acceptable limitinq CPR chanqes for the LRMR and the FMCF events are 0.02 and -0.01 for the ANF fuel, and 0.01 and -0.01 for the GE fuel, respectively.

The application for the core reload amendment for the fifth fuel cycle include a report titled, "MNP-2 Cycle 5 Plant Transient Analysis,"

Revision 1, in which the licensee advised that based on analysis performed for final feedwater temperature reduction for Cycles 3 and 4 as well as Cycle 5 that "delta CPR changes for the LNRB and FMCF transients are conservatively hounded by adding 0.02 to the delta CPR values for these transients at normal feedwater temperature."

By their supplemental letter dated June 1, 1989, (Ref.

12) they proposed these more conservative limits.

Me conclude that the limits are acceptable.

2.3 Thermal-H draulic Stabilit The licensee has performed stability studies for the two limitinq power/

flow conditions (65K power and 45K flow, and 48K power and 27.6X flow).

The calculated decay ratios are 0.52 and 0.49 for cases with and without feedwater temperature reduction at condition of 65K power and 45K flow.

For 48% power and 27.6X flow case, the calculated decay-decreases from 0.84 to 0.70 by inclusion of the effect of reduction in feedwater tempera-ture.

Since the calculated stability ratios for operation by the use of FFTR are bounded by that of cases with the normal feedwater temperature, the staff concludes that the results of the thermal-hydraulic stability analysis are acceptable for the extended fuel Cycle 4 operation with reduction in feedwater temperatures by as much as 65'F.

2.4 Reactor Vessel BeltLine Materials Reactor vessel beltline materials are embrittled when they are irradiated by neutrons from the core.

The procedures for calculatinq the amount of neutron radiation embrittlement of reactor vessel materials are contained in revision 2 to Regulatory Guide (RG) 1.99.

Section 1.3 of the quide

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indicates that the procedures in the quide are applicable for irradiation temperatures between 525'F and 590'F.

The licensee indicated that the minimum water temperature adjacent to the reactor vessel beltline was equivalent to the core inlet temperature, which for WNP-2 during normal power operation was 526.5'F to 533'F.

,The licensee indicated that the FFTR and thermal coastdown would not reduce the water temperature in the beltline region below 525'F.

Hence, the reduction in feedwater tempera-ture would not siqnificantly change the amount of neutron radiation embrittlement and Revision 2 to RG 1.99 may be used to calculate the amount of neutron radiation embrittlement to WNP-2 reactor vessel belt-line materials.

2.5 Reactor Vessel Feedwater Nozzles BWR reactor vessel feedwater nozzles have experienced cracking in the bore and inner radius.

The crackinq was caused by hiqh-cycle and low-cycle thermal fatigue stresses.

The low-cycle thermal fatigue stresses result from system transients.

FFTR will not affect this type of stress.

However, hiqh-cycle thermal fatique stresses will be affected by the FFTR since these stresses result from turbulent mixinq of hot reactor water (approxi-mately 545'F) and the incoming feedwater.

Since the FFTR will increase the difference in temperature between the hot reactor water and the feedwater the high-cycle thermal fatigue stress will increase.

The licensee determined the increase in high-cycle thermal fatique stresses using the rapid cycle duty maps documented by General Electric in topical report NEDE-21821-A (Supplement 2, February 1980).

The staff reviewed the General Electric method of calculatinq high-cycle thermal stresses in NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Crackinq",

November 1980.

The staff concluded that the methods used and the results were acceptable.

The licensee's analysis indicates that the FFTR increased the hiqh-cycle 40-year usage factor from 0.2047 to 0.2796 and the sum of the high-cycle and low-cycle usage factors for 40 years of operation was below the ASf1E Code limit of 1.0.

The increase in usaqe factor assumed that there was no leakage of cold feedwater through the thermal sleeve or its welds. If leakaqe throuqh cracks in either the thermal sleeve or its welds occurred, the licensees analysis would be non-conservative and it is possible that cracks could initiate and propaqate in the feedwater nozzle.

The thermal sleeve and its weld were fabricated usinq austenitic stainless steel material.

Interqranular stress corrosion cracks

( IGSCC) have been observed in austenitic stainless steel materials in BWR reactor water environnments.

NUREG-0313, Rev. 2, January 1900 contains staff recommenda-tions, which should prevent IGSCC of welded austenitic stainless steel material in BWR reactor water environment.

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')3'I 'l' To evaluate whether the thermal sleeve would be susceptible to IGSCC, the staff requested that the licensee identify the materials used to attach the thermal sleeve to the nozzle safe-end.

The thermal sleeve to thermal sleeve extension welds and thermal sleeve extension to nozzle safe-end welds were fabricated using Inconel 182 material.

The thermal sleeves were fabricated usinq 304 stainless steel material with low carbon

(.019%).

The thermal sleeve extensions were fabricated from Inconel 600.

Based on the recommendations in NUREG-0313, Rev. 2, all these materials, except for the Inconel 182 weld metals should not be susceptible to IGSCC.

NUREG-0313, Rev. 2, indicates that Inconel 82 is the only commonly used nickel base weld considered to be resistant to IGSCC.

Since Inconel 182, a nickel base weld metal, was utilized in the thermal sleeve extension welds, the weld must be considered susceptible to IGSCC and leakaqe.

Table 4-31 in NEDE 21821-A, a

GE Tropical Report on this

subject, indicates that if leakaqe increases from 0 GPN to 1.0 GPN throuqh a thermal sleeve, the hiqh cycle fatigue usage factor will increase by a factor of 6 and 225 for 420'F and 340'F rated feedwater temperature, respectively.

This indicates that reducinq the FFTR to 355'F and small amounts of leakage of feedwater through a cracked thermal sleeve during thermal coastdown could substantially increase the fatique usaqe factor for the feedwater nozzle.

Hence, leakage through cracks in the Inconel 182 weld metal and reducinq the FFTR to 355'F could affect the inteqrity of the feedwater nozzle.

By letter dated February 14, 1990 (Ref. 14) the licensee indicated that the feedwater nozzle inspection proqram would be revised to schedule a

volumetric inspection (ultransonic technique) of at least one nozzle during the refuelinq outage commencing after each use of feedwater temperature reduction to extend the fuel cycle.

With this auqmented inspection program to ensure that leakage throuqh the Inconel 182 weld metal in the feedwater thermal sleeve extension welds does not result in crackinq in the feedwater

nozzle, the potential impact to the feedwater nozzles is acceptable.

2.6 Reactor Internals Load Im act All the reactor internals (e.q.,

core plate, shroud support,

shroud, top

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head, steam dryer, control rod quide tube, control rod drive housing and jet pump) were evaluated under the consideration of additional loads imposed by the ICF and FFWTR operations.

Based on the duced in those components are within the ASNE Code,Section III, Subsection NG allowables and the criteria referenced in the FSAR.

The staff finds this acceptable since the original desiqn criteria were satisfied.

2.7 Flow-Induced Vibration To ensure that the flow-induced vibration response of the reactor internals for plant operation with ICF up to 106% rated flow is acceptable, the prototype (Tnkai 2) plant test data as well as data from WNP-2 testing up to 106% core flow were reviewed.

WNP-2 reactor internals were tested in accordance with provisions of Regulatory Guide 1.20, Revision 2 for

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non-prototype, Category IV plants usinq Tokai-2 as the limited valid pro-totype.

This was described in the FSAR and was previously accepted by the staff.

Based on the results of the licensee's assessment, the maximum alternatinq stress intensity for core flow up to 106% is 92% of the FSAR reactor internal vibration acceptance criterion.

Therefore, the operation of WNP-2 at 106% rated core flow will not result in unacceptable reactor internal vibration.

2.8 Feedwater Nozzle and Feedwater S arqer Fatigue Usaqe At the end of the 1970's, inspections at 22 boi ling water reactor plants identified crackinq in the feedwater nozzle and sparger at 18 reactor vessels.

The staff studied the issue and recommended hardware modifica-tions, analysis methods and inspection schedules for the.nozzle and sparger in NUREG-0619.

The proposed FFWTR requested by the licensee will affect the fatique usage of the feedwater nozzle and sparger.

The licensee indicated that the increase of'he fatigue usaqe factor due to feedwater temperature reduction was calculated using the analysis method described in GE report NEDE-21821-A (Supplement 2, February 1980).

This GE report was referred to in NUREG-0619 and was approved by the staff in a letter from D.

G. Eisenhut to R.

G. Gridley dated January 14, 1980.

Based on the result of the licensee's

analysis, the 40-year total fatique usaqe factor with the proposed FFWTR operation remains below the ASt1E Code limit of 1.0 and is thus considered acceptable.

3.0 Technical S ecification Chanqes The followinq changes to the Technical Specifications and Bases, have been proposed for operation to extend the fuel cycle by reducinq feedwater temperatures.

3.]

Technical S ecification Sections 1.12A and 1.13A One definition is added to each of the sections.

The definitions are for the End-nf-Cycle (EOC) and final feedwater temperature reduction, respect-ively.

The changes are necessary to specify the conditions for the extended fuel cycle operation and are acceptable.

3.2 Technical S ecifications Section 3/4.1.6 This is a new section, the limitinq condition of operation (LCO) and surveillance requirements are added for the extended cycle operation with feedwater temperature reduction below 355'F.

The changes are consistent with the current Technical Specifications with feedwater temperature reduced below the normal feedwater temperature of 420'F and are acceptable.

3.3 Technical S ecifications Table 3.2.3.1 The operatinq NCPR limits based on the measured control rod scram times as well as the NCPR limits based on the control rod insertion time bounded by

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The chanqes are consistent with analyt-ical results and are acceptable.

3.4 Figure 3.2.3-1 The note below the fiqure is revised to remove the phrase "when approved" since the issuance of this amendment constitutes approval of final feedwater temperature reduction.

The curve is supported by analytical results and is acceptable.

3.5 Technical S ecifications Basis 3/4.1.6.

A section is added to specify the feedwater temperature during the extended fuel cycle operation.

The chanqe is consistent with the assumptions used for the analysis to support the request of the Technical Specification chanqes and is acceptable.

3.6 Technical Specifications Basis 3/4.2-3 The statement is added to this section, which identifies the load reiection without bypass as the limiting case used to determine the operatinq MCPR limits for the extended fuel cycle.

This chanqe is supported by the analytical results discussed in this evaluation and is acceptable.

4.0 NO SIGNIFICANT HAZARDS DETERMINATION The Commission has provided standards for determining whether a siqnifi-cant hazards consideration exists as stated in i0 CFR 50.92.

A proposed amendment to an operating license for a faci lity involves no siqnificant hazards considerations if operation of the facility in accordance with a proposed amendment would not:

(1) Involve a siqnificant increase in the probability or consequences of an accident previously evaluated; (2)

Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

The proposed amendment does not represent a siqnificant hazard because it does not:

1)

Involve a significant increase in the probability or consequences of an accident previously evaluated.

The amendment allows operation with reduced feedwater temperature at the end of the fuel cycle to extend the length of the fuel cycle.

The licensee analyzed the safety limits for the existing and reload fuel for the last three reloads and established safety limits to be applicable for the period during which feedwater temperature is reduced.

The staff review of the licensee's submittal concluded that the analyses performed by the licensee and the limits established are acceptable.

The licensee also analyzed the impact of the proposed operation with reduced feedwater temperature on reactor vessel beltline material,

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on thermal stresses to feedwater nozzles, on mechanical stresses to reactor internals, on flow induced vibrations, and on material failure due to feedwater nozzle and feedwater sparqer fatigue usage.

The staff review found no significant change in previously accepted conditions.

2)

Create the possibi lity of a new or different kind of accident from any accident previously evaluated.

The proposed chanqe does not involve installation of new or different components.

The amendment is limited to changes in operatinq procedures.

Furthermore, the operational limits applied to WNP-2 have not been changed.

Values for these limits to be applied during final feedwater temperature reduction were derived from NRC qualified computer codes and by applying the most limitinq transients throughout the cycle.

The review of the proposed limits and of the impact on mechanical com-ponents and structures did not identify the potential for any new or different kind of accident.

3)

Involve a significant reduction in a marqin of safety.

The licensee's analysis of safety limits to be applicable during periods of reduced feedwater temperature resulted in a margin of safety either identical to or more conservative than that now used for the plant.

The staff found the methods used by the licensee acceptable.

The staff also completed its review of potential impacts to mechanical structures due to chanqes in core flow durinq feedwater temperature reduction period and metallurgical stresses due to reduced feedwater tempera-ture and found these impacts acceptable.

Therefore the stai'f con-cludes the change does not involve a significant reduction in a margin of safety.

For these reasons the staff has determined that the proposed amendment involves no siqnificant hazards consideration.

5.0 ENVIRONMENTAL CONSIDERATION

This amendment involves a chanqe in a requirement with respect to the installation and use of a facility component located within the re-stricted area as defined in 10 CFR Part 20 and changes surveillance requirements.

The staff has determined that this amendment involves no siqnificant increase in the amounts, and no siqnificant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radia-tion exposure.

The staff has determined that this amendment involves no significant hazards consideration.

Accordinqly, this amendment meets the eliqibility criteria for categorical exclusion set forth in 10

CFP, 51.22(c)(9).

Pursuant to 10 CFR 51.22(b),

no environmental impact state-ment or environmental assessment need be prepared in connection with the issuance of this amendment.

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6.0 CONTACT WITH STATE OFFICIAL The Commission published a notice of consideration of issuance of amendment to facility license and opportunity for hearinq (53 FR 7270, March 7, 1988).

No request for hearing or petition for leave to inter-vene was received.

By letter dated March 31, 1989, the State of Washinqton advised that they do not have any comment.

7. 0 CONCLUSION In our review of the proposed changes to the Technical Specifications for operation of Cycle 4 with final feedwater temperature reduction, we find that approved methods have been

used, and the results of the feedwater temperature reduction analysis support the proposed operating MCPR limits, which avoid violation of the safety MCPR limit for desiqn transients.

Therefore, the staff concludes that this core design will not adversely affect the capability to operate the plant safely durinq the extended fuel Cycle 4 operation at lower feedwater temperatures and the proposed Technical Specification chanqes are acceptable.

The calculated CPR chanqes due to the effect of FFTR are also acceptable for the future reload appli-cations provided that the assumptions used in performinq transient analysis remain consistent with the assumptions used in Reference 4.

The method of calculatinq neutron radiation damaqe to reactor vessel belt-line materials documented in revision 2 to RG 1.99 will be applicable for WNP-2 when the FFTR temperature is reduced from 420'F to 355'F and during thermal coastdown.

Since the materials used to fabricate the thermal sleeves and thermal sleeve extensions are not susceptible to IGSCC, leakage of cold feedwater throuqh the thermal sleeve is not likely.

Hence, the method used by the licensee to calculate hiqh-cycle thermal fatique stresses is acceptable and the usaqe factor for the feedwater nozzles during the 40 year life of the Plant should not exceed the ASME Code limit of 1.0.

The licensee has provided an adequate justification with respect to load impact of reactor internals, flow-induced vibrations as well as feedwater nozzle and sparqer fatique usage factor for the operation of WNP-2 with FFWTR and ICF up to 1064 rated flow.

Based nn the above conclusions, FFTR and thermal coastdown will not affect reactor vessel material,integrity and the requested chanqes in technical specifications should be approved.

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 endangered by 'operation in the proposed

manner, and (2) such activities will be conducted in compliance with the Commission s requla-tions and (3) the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

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8.0 RFFERENCES Letter from G. Sorensen (WPSS) to NRC, dated December 15, 1987 (G02-87-286)..

2.

3.

6.

7.

9.

NEDC-31107, Safety Review of WPPSS Nuclear Project No.

2 at Core Flow Conditions Above Rated Flow Throuqhout Cycle 1 and Fuel Feedwater Temperature Reductions, March 1986.

WPPSS-EANF-111, Final Feedwater Temperature Reduction Summary

Report, November 1987.

XN-NF-87-92, WNP-2 Plant Transient Analysis with Final Feedwater Temperature Reduction, June 1987.

XN-NF-79-71, Revision 2 (as supplemented),

Exxon Nuclear Plant Transient Methodoloqy for Boilinq Water Reactors, February 1987.

XN-NF-105(A), XCOBRA-T:

A Computer Code for BWR Transient Thermal-Hydraulic Code Analysis, February 1987.

Letter from G. Sorensen (MPPSS) to NRC, dated September 14, 1988 (G02-88-198) and XN-NF-87-92 (Supplement 1), MNP-2 Plant Transient Analysis with Final Feedwater Temperature Reduction Cycle 4 Analysis, Hay 1988.

Letter from G. Sorensen (WPPSS) to NRC, dated Yiarch 3, 1989 (G02-89-029).

Letter from G. Sorensen (WPPSS) to NRC, dated March 7, 1988 (G02-88-054).

10.

Letter from G. Sorensen (WPPSS) to VRC, dated April 12, 1988 (G02-88-087).

11.

Letter from G. Sorensen (MPPSS) to NRC, dated April PO, 1989 (G02-89-067).

12.

Letter from G. Sorensen (MPPSS) to NRC, dated June 1, 1989 (G02-89-102).

13.

Letter from G. Sorensen (MPPSS) to NRC dated February 14, 1990 (G02-90-024).

Principal Contributors:

S.

B. Sun Barry J. Elliot Renee Li Dated.

March 1, 1990

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