ML060950413

SER Open Item 4.7B.1-1 and Closure of Amended License Renewal Application (Alra) Section A2.4, Commitment 39 - Biological Shield Wall Neutron Fluence Analysis

ML060950413
Person / Time
Site:	Nine Mile Point
Issue date:	03/23/2006
From:	O'Connor T Constellation Energy Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC MC3272, TAC MC3273
Download: ML060950413 (9)

Text

.. Constellation Energy Nine Mile Point Nuclear Station P.O. Box 63 Lycoming, NY 13093 March 23, 2006 U. S. Niclear Regulatory Commission Washington, DC 20555-0001 ATTENTION:

SUBJECT:

Document Control Desk Nine Mile Point Nuclear Station Unit Nos. 1 & 2; Docket Nos. 50-220 & 50-410 SER Open Item 4.7B.1-1 and Closure of Amended License Renewal Application (ALRA) Section A2.4, Commitment 39 -Nine Mile Point Unit 2 (NMP2) Biological Shield Wall Neutron Fluence Analysis (TAC Nos. MC3272 and MC3273)By letter dated January 11, 2006, Nine Mile Point Nuclear Station, LLC (NMPNS) submitted a response to NRC aRAI 4.7.1 B-1 in which it committed to submit, for NRC review and approval, the summary of the Reg. Guide 1.190 based analysis that determines the maximum neutron fluence at the NMP2 Biological Shield Wall or at the shield wall flaw locations that were the basis for the ALRA Section 4.7.1 Time-Limited Aging Analysis (TLAA). This analysis is addressed as Open Item 4.7B. 1-1 in the NRC's"Safety Evaluation Report With Open Items Related to the License Renewal of Nine Mile Point Nuclear Station, Units 1 and 2", dated March 2006.The subject analysis has been completed.

Attachment 1 provides a summary of the results of the analysis and provides the revisions to the NMPNS ALRA that are required based on those results. This letter contains no new regulatory commitments.

Should you have questions regarding the information in this submittal, please contact P. A. Mazzaferro, NMPNS License Renewal Project Manager, at (315) 349-1019.Nine Mile Point A I (D r7 Document Control Desk March .23, 2006 Page 2 STATE OF NEW YORK TO WIT: COUNTY OF OSWEGO I, Timothy J. O'Connor, begin duly sworn, state that I am Vice President Nine Mile Point, and that I am duly authorized to execute and file this submittal on behalf of Nine Mile Point Nuclear Station, LLC. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other Nine Mile Point employees and/or consultants.

Such information has been reviewed in accordance with company practice and I believe it to be reliable.Subscribed and sworn before me, a Notary Public in and for the tate of New York and County of Oswego, this .2 3 day of i 2006.WITNESS my Hand and Notarial Seal: A 7 §L Q Notary Public My Commission Expires: 3/23/a 6 Date Cot TJO/MRZF/sac Attachments:

(1) SER Open Item 4.7B.1-1 and Closure of ALRA Section A2.4, Commitment 39 -NMP2 Biological Shield Wall Neutron Fluence Analysis cc: S. J. Collins, NRC T. G. Colburn, NRC N. B. Lee, NRC Resident Inspector, NRC J. P. Spath, NYSERDA ATTACHMENT (1)SER OPEN ITEM 4.7B.1-1 AND CLOSURE OF ALRA SECTION A2.4, COM.%IMITMENT 39 -NMP2 BIOLOGICAL SHIELD WALL NEUTRON FLUENCE ANALYSIS Nine Mile Point Nuclear Station, LLC March 23, 2006 ATTACHMENT (1)SER OPEN ITEM 4.7B.1-1 AND CLOSURE OF ALRA SECTION A2.4, COMMITMENT

39) -NMP2 BIOLOGICAL SOIELD WALL NEUTRON FLUENCE ANALYSIS In NMPNS Letter NMP1L 2015, dated January 11, 2006, in response to NRC aRAI 4.7. 1B-1, the following commitment was made: "No later than two years prior to entry in the PEO, NMP will submit, for NRC review and approval, the summary of the Reg. Guide 1.190 based analysis that determines the maximum neutron fluence at the NMP2 Biological Shield Wall or at the shield wall flaw locations that are the basis for the ALRA Section 4.7.1 TLAA. The submittal will include revised ALRA Sections 4.7.1 and A2.2.5.1, and any other supporting analysis, as applicable." This commitment was made because the referenced TLAA (ALRA Section 4.7.1), on neutron embritilement of the NMP2 Biological Shield Wall (BSW), was based on the initial fluence analysis that was performed for NMP2 that used the fluence analysis methodology that was accepted by the NRC at that time. Using that methodology, the calculated neutron fluence at the BSW was greater than the 10 CFR 50, Appendix H defined threshold of IE17 n/cm2 for the Onset of neutron embrittlement of steel. Based on that calculated fluence and the presence of mino: indications in the BSW, a fracture mechanics analysis of the BSW was performed that qualified as a TLAA under 10 CFR 54.3(a).Since that time, Reg. Guide 1.190 has been issued for the performance of fluence analyses and the methodology in this guide is less conservative that the methodology utilized in the original fluence analysis.

The Reg. Guide 1.190 methodology was utilized in the NRC approved reactor pressure vessel neutron embrittlement analyses addressed in NMPNS ALRA Section 4.2. These analyses and the fluence analysis methodology utilized were approved by the NRC in its SER transmitted via letter to NMPNS dated October 27, 2003 (TAC No. MB6687). The same fluence analysis model utilized in the reactor vessel neutron embrittlement analyses was re-run with receptor points for the BSW to determine the neutron fluence at the inside of the wall.The results of this reanalysis are presented as follows to fulfill the above commitment.

ANALYSIS

SUMMARY

The revised NMP2 BSW fluence evaluation is defined using the NMP2 transport calculation methods approved for use for NMP2 reactor vessel fluence applications.

The NRC reviewed and accepted these methods in the "Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 183 to Facility Operating License No. Dpr-63 Nine Mile Point Nuclear Station, LLC Nine Mile Point Nuclear Station, Unit No. 1 Docket No. 50-220." The review determined that these methods met the requirements of RG 1.190 and the Safety Evaluation concluded, in the section entitled "Acceptability of Neutron Transport Calculation": "Bssed on review of the licensee's submittals set forth above, the NRC staff finds that the licensee used acceptable methodology to derive the applicable fluence values. The licensee's September 15, 2003, letter submitted the revised version of the benchmarking report ("Benchmarking of Nine Mile Point Unit 1 and Unit 2 Neutron Transport Calculations," MPM-402781 (Revision 1), September, 2003, MPM Technologies, Inc.) to update the dala, discussions, and conclusions contained in the licensee's July 31, 2003 letter." 1 of6

1 ATTACHMENT (1)SER OPEN ITEM 4.7B.1-1 AND CLOSURE OF ALRA SECTION A2.4, COMMITMENT

39) -NMP2 BIOLOGICAL SHIELD WALL NEUTRON FLUENCE ANALYSIS The calculations in the active fuel region were carried out using a synthesis of two dimensional neutron transport calculations, including plant-specific R-0 and R-Z calculations, for each fuel cycle through the end of NMP2 Cycle 10 (14.08 EFPY, current operating cycle) as described in this benchmark report. Each active fuel region case consisted of three transport analyses (R-17, R-Z, and R), which were synthesized to provide a three dimensional flux profile at the shroud, vessel, and BSW. The active fuel region transport synthesis includes the air gap and the biological shield wall. The calculations for the BSW are defined for end of Cycle 10 and project the fluence for this location to exposures of 22 EFPY and 54 EFPY (end of the period of extended operation).

The beltline region calculation procedures, which include the BSW, meet standards specified by the NRC and ASTM, as appropriate, and meet the requirements of Reg.Guide 1.190.At the inside of the concrete shield is a steel shield plate 1.5 inches in thickness.

The maximum fluence to this plate was evaluated similarly to the analysis of the reactor vessel in the beltline region. The maximum BSW fluence occurs at the plate inner radius and the values at this rat ius were determined as a function of azimuthal angle and axial height. The variation with angle i s tempered by streaming within the cavity and the greater distance from the core. The maximum fluence was found to be at an angle of about 30 degrees in the first octant. The azimuthal distribution varies slightly with each cycle. The maximum fluence to the shield at the end of Cycle 10 was found to be less than 2E16 n/cm 2 and the maximum fluence at 54 EFPY was found to be 6.2E16 n/cm 2 for fast neutrons (E>1.0 MeV).Since the 54 EFPY maximum fluence value at the BSW is less than the threshold fluence value (1E17 rI/cm 2) for the susceptibility of steel to neutron embrittlement identified in 10 CFR 50, Appendix H, the consideration of this aging effect no longer applies. Since the aging effect no longer applies, TLAA criterion 10 CFR 54.3(a)(2) is no longer applicable to the original analysis.

Since all six (6) of the 10 CFR 54.3(a) criteria must apply for an analysis to qualify as a TLAA, the analysis described in ALRA Section 4.7.1 no longer qualifies as a TLAA; therefore, the ALRA needs to be revised accordingly.

ALRA REVISIONS With the elimination of the NMP2 BSW TLAA, there are several revisions to the ALRA that are shown beginning on the following page. Additions to the ALRA are shown with italicized font and deletions are shown with text strikethreughs.

2of6 ATTACHMENT (1)SER OPEN ITEM 4.7B.1-1 AND CLOSURE OF ALRA SECTION A2.4, COMMITMENT 39 -NMP2 BIOLOGICAL SHIELD WALL NEUTRON FLUENCE ANALYSIS On page 4.1-3 of the ALRA, Table 4.1-1 is revised as shown below.Table 4.1-1 Time-Limited Aging Analyses Applicable to NMPNS CategcLAy Description Disposyion Section 1. Reactor Vessel Neutron Embrittlement Analysis 4.2 Upper-shelf Energy §54.21 (c)(1)(ii) 4.2.1 Pressure-Temperature (P-T) Limits §54.21 (c)(1)(iii) 4.2.2 Elimination of Circumferential Weld Inspection (NMP1 only) §54.21 (c)(1)(ii) 42.3_ Axial Weld Failure Probability

§54.21 (c)(1)(ii) 4.2.4 2. Metal Fatigue Analysis 4.3 Reactor Vessel Fatigue Analysis §54.21 (c)(1)(iii) 4.3.1 ASME Section III Class 1 Piping and Components Fatigue §54.21(c)(1)(iii) 4_3.2 Analysis (NMP2 only)Feedwater (FWS) Nozzle and Control Rod Drive Return Line §54.21 (c)(1)(iii) 4.3.3 (CRDRL) Nozzle Fatigue and Cracking Analyses Non-ASME Section III Class I Piping and Components

§54.21(c)(1)(iii) 4_3.4 Fatigue Analysis Reactor Vessel Internals Fatigue Analysis §54.21 (c)(1)(iii) 4.3.5 Environmentally Assisted Fatigue §54.21 (c)(1)(iii) 4.3.6 Fatigue of the Emergency Condenser (NMP1 only) §54.21 (c)(1)(iii) 4.3.7 3. Environmental Qualification (EQ) 4 4 Electrical Equipment EQ §54.21 (c)(1)(iii) 4.4.1 Mechanical Equipment EQ (NMP2 only) §54.21 (c)(1)(iii) 4.4.2 4. Concrete Containment Tendon Prestress Analysis Not Applicable 4 5 5. Containment Liner Plate, Metal Containments, and 46_ Penetrations Fatigue Analysis Torus Shell and Vent System Fatigue Analysis (NMP1 only) §54.21 (c)(1)(i) 4.6.1 and§54.21 (c)(1)(ii)

Torus Attached Piping Analysis (NMP1 only) §54.21 (c)(1)(iii) 4.6.2 Torus Wall Thickness (NMP1 only) §54.21 (c)(1)(iii) 4.6.3 Containment Liner Analysis (NMP2 only) §54.21 (c)(1)(ii) 4.6.4 Fatigue of Primary Containment Penetrations

§54.21 (c)(1)(i), 4.6.5§54.21 (c)(1)(ii) and§54.21 (c)(1)(iii)

Downcomer and Safety/Relief Valve Discharge Line Fatigue §54.21 (c)(1)(ii) 4.6.6 Evaluation (NMP2) Only and_ __ §54.21 (c)(1)(iii)

6. Other Plant-specific TLAAs 4.7 RPV Biological Shiold (NMP2-only)Deleted j2 j( 4.,.1 Main Steam Isolation Valve Corrosion Allowance

§54.21(c)(1)(iii) 4.7.2 (NMP2 only)Stress Relaxation of Core Plate Hold-Down Bolts §54.21 (c)(1)(iii) 4'.3 (NMP2 only)Reactor Vessel and Reactor Vessel Closure Head Weld Flaw §54.21 (c)(1)(i)

4. .4 Evaluations (NMP1 only) and__ §54.21 (c)(1)(iii) 3 of6 ATTACHMENT (1)SER OPEN ITEM 4.7B.l-1 AND CLOSURE OF ALRA SECTION A2.4, COMMITMENT 3 9 -NMP2 BIOLOGICAL SHIELD WALL NEUTRON FLUENCE ANALYSIS ALRA Section 4.7.1, on pages 4.7-1, -2, and -3, is revised as shown below.4.7 OTHER PLANT-SPECIFIC TLAAS 4.7.1 RPI VBIOLOGICAL SHI1ELD (NAIP2 ONLY)DELETED Summqire Dcscription, A biological shield wall (BSW) with an inner radius of 14 feet, 3/ inch and an outer radius of 15 feet, 9 t/ inch surrounds the N4P2 RPNV. The BSW consists oftho concentric 1 / inch thiek steel cylinders connected by internal h otal and veriea;: stiffeners.

Full penetration welds conct the plates that make up the elinders.

The space between the steel cylinders is filled mith nonstructural heavy density fill matcria for radiation shielding. (Refer to Section 3.8.3.1.3 of Reference 4.8 34.)Discovery of weld defeats during fabrcation of the BSW resulted in strcss and fractu e me s analyses to detenrine an acceptable flaw size; the results showed the m riy of the flaws were aceeptable, while a small number of flaws required repair (Tabls 1'-2 and 3 in enclosurc to Refcr-ence 4.8 65). A related calculation was prepared to estim .4 the amount of neutron irradiation embrittlement (in terms of the 30 ft lb transition temperature shift) of the BSW structural steel at the end of a 40 year life. Since this calculation confirmed thc validity of the BSW fracture mechanics analyses for the cui~ent license term, it satisfies the criteria of §54.3(a).

As such, this analysis is a TLAA.A threshold fluene value was determined belowA which the transition temperature shif would be zero. The 10 year neutron fluence at the BSW inside surface was determin:

e" be less than the threshold value; therefore, the conclusion of the subject calculation staes that no neutron embr-ittlement of-the stfuctur-al steel weuld occur during. the 10 year lifie ef the plant.The original fracture mechanics analysis specified that the stress intensity factor (Ka'be less than a dynamic fracture toughness ( ) of 48.8 ksi(in)", based on a Char-py V nteh enefgy (C-,) of 20 ft lbs at 100F. This value was applied as an aceeptance criterion fr flaws in the base metal. Measured Cv values shcwed that the K values used for the weld metal and heat ginafcdne in the al calculation both had highe fracture toughncss than the base metal. Since the shift in the Gvalues for the weld and heat affected zone was expected to be no greater than that predicted for the base metal, the basc metal toughness was eensider-ed bounding for this cvaluation (Section IW.B.2 in enllrsure te Referenee 4 The NRC reviewed the repairs to the BS3A w, elds and the associated fracture mechanies evaluations, and concluded that all BSW welds were acceptable for the intended serviee (Section 2.2 in Enclosure to Reference 4.8 66).4of6 ATTACHMENT (1)SER OPEN ITEM 4.7B.1-1 AND CLOSURE OF ALRA SECTION A2.4, COMMITMENT 39 -NMP2 BIOLOGICAL SHIELD WALL NEUTRON FLUENCE ANALYSIS Di^Asiresiten RCA4.21 *e)1)(ii)

T-heanalysesh^,ebepfjetd heend. +Afh reA+period of extended operation.

At the BSW outer- wall, neutroen fluence is negliggible due to attenution thfough the heavy density fill material; therefore, the fracture toughess properties of the outer w d4 platcs and welds will be unaffected.

The neutron fluenee at the surface of the BSW imer -wall has been projeeted tharough Te period of extended operation.

For E>1.0 MleV, the most recent RPV sun eillance repc4 (the attachment to Reference 4.8 20) documents a prejected peak fluenee at the RPV inSne r radius Cof 5r1 1n/n nat. 8.722 EPY, with an average flux value of 8.7xlO 8-nlem s -at the same location.

This flux value can be used to extrapolate the fluence for an additional 45.28 EFPY exposure, yielding a fluenee value of I .4!9xlO+_8-n/cmat 54 EFPY. A conservative value of the corresponding fluence at the RPNZ outer radius is predicted-byf multipling the inner surface fluence value by the exponential attenuation facter(ei'-

is the thie lmessof the RPYwall (6.41375 inhes, determined from Table 3 2 in the attachment to Reference 4.8 20). The neutro flux with E>1.0 Men falls off by approximately 18%' in the void between the exterior surfaee of the RPV and the BSW inside surface; thus, a 54 EFPY fluence of 2.54x10 4-mem is projected at the surface ef the inner steel cylinder of the BS).More recent data fcr irradiation of structural steels at low temperatures enables a more accurate estimation of embrifflement for- tlle BS)A1. Materials from the Shippingport Reactor neutron shield tank and the High Flux Isotope Reactor vessel vvere irradiad t 507xlO"-nkm E-2 1 Men ) in a test r-eactor at a controlled temperature of 130F to+approeximate the ncal service temperattires f the statures.

The results indicatedt elevation in 30 lb trasition temper-ature of 35F and a rCeduction in USE ef less than 6 ft lb (RPference 4.8 67). Since the projected fluenee for the NMP2 BSW is less than the value reported in Refer nee 4.8 67, thc shift in C. due to irradiation is alse reduced. Reduction in material properties due to irradiation has been showin to be proportional to the square root of fluence for low fluenee irradiation; thus, the reductien inCM, energy at 100°F wvas determined by multiplying the 30 f+ lb temperature shift at 07Ow mtrA271repor+

ed in Reference 4.8 67) by the ratio of the square roots of the projected fluenee at the BSW inner- wall and the r-efefence fluene (07x1 n This results in a revised C _ for the BSW steel of 9.62 f+ lbs at 1 OT, and K +ef 37.1 ksi(inf)".

A review of Tables 1, 2, and 3 in the enclosure to Reference 4.8 65 shows that no indications with applied greater than or equal to the projected K were allowed to remain in service with Bsed on projected fluenee value, the USE of the BSW material is reduced but does not invalidate the original fracture mechanics analyses.Therefore, fracturc toug.hncss of the NA92 BSWM has been pr-ejeeted (reev aluated)-fer-th period of extended operation in accordance with §54.21(e)(1)(ii).This section deleted.5 of 6 ATTACHMENT (1)SER OPEN ITEM 4.7B.1-1 AND CLOSURE OF ALRA SECTION A2.4, COMMITMENT

39) -NMP2 BIOLOGICAL SHIELD WALL NEUTRON FLUENCE ANALYSIS On ALRA pages 4.8-7 and 4.8-8, References 4.8-65, -66, and -67 are deleted as shown below.4.8-65 Letter from Niagara Mohawkl Power Corporation to U.S. Nuclear Regulatory Commission dated August 1, 1980 fonvarding the final repoA eeneemi g the Nine Mile Point Unit 2 biological shield wall in a*VIrdaree with 10 CFR 50, paragnaph 50.55(e)(3).(deleted)

Letter from U.S. Nuclear Regulatory Commission to Niagara MIohawA 3nwer Cnnrniati'n clata '9numh 1r 45)5 I U i,9t T43=010tiW3 MO UL 4.8-66 I. -II -1 I. -. -.- -, -, ___.1 __.. II.1W__.._..

I- --4ja43:5 29.(deleted) 4.8-67 SAND92 2420, MEA 2494, ,4ecc-rated 54°C-Irradiated Tcst of Shzppingport NCUX On Shicd Tank. and lIUR Vesscl aMW&ials, January 1993(deleted)

On ALRA page A2-30, Section A2.2.5.1 is revised as shown below.A2.2.51 RPY BIOLOGICAL SHIELDDELETED Discovery of weld defeAts during abcation of the Bielogieal Shield Wall (BSWl resulted in stress and fracture mechanics analyses to determine an aceeptable flaw size. The results showed the majority of the flaws were acceptable, while some flaws required repair (Referenee A2.3.4). A related calculation was prepared to estimate the amount of neutron irradiation embrittlement (in terms of the 30 f lb transitio temperature shift) of the BSW structural steel at the end of a 40 year life.Based on projected fluence value, the USE cf the BSW material is reduced but does not invalidate the original fracture mechanics analyses.

Therefore, fracture toug wess of the bTh4P2 BSW has been projected (reevaluated) for the period of extende eperatien.YThis section deleted.6of6