ML120470088

Supplement to 10CFR50.55a Requests Associated with the Fourth and Fifth Lnservice Lnspection Ten-Year Intervals

ML120470088
Person / Time
Site:	Monticello
Issue date:	02/15/2012
From:	O'Connor T J Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-12-011
Download: ML120470088 (16)

Text

@ Xcel Energya February 15,2012 Monticello Nuclear Generating Plant 2807 W County Road 75 Monticello, MN 55362 L-MT-12-011 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket 50-263 Renewed Facility Operating License No. DPR-22 Supplement to 1 OCFR50.55a Requests Associated with the Fourth and Fifth lnservice lnspection Ten-Year Intervals

References:

1) BWRVIP-108, "BWR Vessel and Internal Project Technical Basis for the Reduction of lnspection Requirements for Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," dated October 2002. 2) NSPM to NRC, "10 CFR 50.55a Request No. 16: Alternative to Nozzle-to-Vessel Weld and lnner Radius Examinations," dated March 12,201 0. (Accession No. ML100750660)

3) NRC to NSPM, "Monticello Nuclear Generating Plant - Alternative to Nozzle-to-Vessel Weld and lnner Radius Examinations, (TAC No. ME3527)," dated November 24, 201 0. (Accession No. MLI 031 9031 1) 4) NSPM to NRC, "10 CFR 50.55a Requests Associated with the Fifth lnservice lnspection Ten-Year Interval," dated September 28, 201 1. (Accession No.

MLI 12720147)

5) Structural Integrity Associates, Inc., Calculation Package, Monticello N2 Nozzle Code Case N-702 Relief Request, File No. 1101463.301 Northern States Power Company, a Minnesota corporation, d/b/a Xcel Energy (hereafter "NSPM"), has identified that Boiling Water Reactor Vessel lnternals Project (BWRVIP) report, BWRVIP-108 (Reference I), contains certain design assumptions that impact its application at the Monticello Nuclear Generating Plant (MNGP). Specifically, BWRVIP-108 assumed a certain number of thermal cycles on the reactor pressure vessel (RPV) nozzles assuming a 40 year design life. Also, the report assumed that the Recirculation Inlet (N2) nozzles accumulated negligible fluence allowing them to be excluded them from the "beltline" region of the RPV. The L-MT-12-011 Document Control Desk Page 2 of 3 BWRVIP-108 report served as part of the basis for two 10 CFR 50.55a requests (References 2 and 4) to implement an alternative, American Society of Mechanical Engineers (ASME) Code Case N-702.

Subsequently, to support use of ASME Code Case N-702, Structural Integrity Associates, Inc. (Reference 5), performed a MNGP site specific analysis for NSPM that evaluates I) additional thermal cycles assumed during the 60 year design life following approval of license renewal, and

2) additional fluence at the "beltline" region for the RPV N2 nozzles. Enclosure 1 provides this analysis which demonstrates the acceptability of applying ASME Code Case N-702 through the end of the renewed license period of extended operation.

NSPM requests that the U.S. Nuclear Regulatory Commission (NRC) review this site specific analysis together with the pending "1 0 CFR 50.55a Requests Associated with the Fifth lnservice lnspection Ten-Year Interval," dated September 28, 201 1 (Reference 4), specifically as it pertains to 10 CFR 50.55a Request RR-002, "Alternative to the Requirements of Examination Category B-D." On March 12, 2010, NSPM submitted 10 CFR 50.55a Request No. 16, (Reference 2), pursuant to 10 CFR 50.55a(a)(3)(i) in accordance with ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds." Part of the basis for this 10 CFR 50.55a request was BWRVIP-108. The NRC reviewed this 10 CFR 50.55a request and authorized this alternative for application to MNGP on November 24, 2010, through the end of the Fourth Ten-Year lnservice lnspection (ISI) Interval (Reference 3).

NSPM requests that the NRC review the site specific analysis provided herein, and pursuant to 10 CFR 50.55a(a)(3)(i), reauthorize application of 10 CFR 50.55a Request No. 16 through the Fourth Ten-Year IS1 Interval. Reauthorization is necessary to allow use of the ASME Code Case N-702 alternative for the duration of the Fourth Interval, to maintain MNGP compliance with 10 CFR 50.55a(a)(3)(i).

NSPM thereby requests that the NRC complete reauthorization prior to the end of the MNGP Fourth Ten-Year IS1 Interval, scheduled for August 31, 2012.

Should you have questions regarding this letter, please contact Mr. Randy Rippy at (61 2) 330-691 1. ommitments and no revisions to existing commitments. resident, Monticello Nuclear Generating Plant

~ortheph States Power Company - Minnesota L-MT-12-01 I Document Control Desk Page 3 of 3 Enclosure cc: Administrator, Region Ill, USNRC Resident Inspector, Monticello, USNRC Project Manager, Monticello, USNRC BWRVIP Project Manager, USNRC ENCLOSURE 1 STRUCTURAL INTEGRITY ASSOCIATES, INC. CALCULATION PACKAGE MONTICELLO N2 NOZZLE CODE CASE N-702 RELIEF REQUEST FILE NO.: 1101463.301 12 pages follow

! . . ! I f' i t' Structural Integrity Associates, Inc. CALCULATION PACKAGE PROJECT NAME: Monticello N2 Nozzle Code Case N-702 Relief'Request CONTRACT NO.: 00001005 Release 0032 CLIENT: PLANT: File No.: 1101463.301 Project No.: 1101463 Quality Program: /ZINuolear 0 Commeroial XCELEnergy Monticello Nuclear Generating Plant CALCULATION TITLE: Evaluation of effect ofillspection on the probability offailu1'8 for BWR Recirculation Iniet (N2) shell-welds and nozzle blend radii region at Monticello Document Revision o Affecte(l Pages 1-10 A-I-A-2 Revision Descl'iption Initial Issue Project Managel' . Approval Signature

& Date Jim Wu Pt'elhll'el'(s)

& Checket*(s)

Signatures

& Date fi., , . Stan Ta SST

/'2. . Page 1 oflO F0306*01RI 1ntesrirv Associates.

he?' Table of Contents 1. 0 OBJECTWE .........................................

..,..... ...............................................................

3 ....................................................................................................

2.0 METHODOLOGY

3 3. 0 DESIGN mW ............................................................................................................

3 4.0 ASSUMPTIONS

...................

.,, .....................................................................................

4 5.0 SOFTWARE MODIFICATIONS

...........................................

4 6.0 FATIGUE GRACK ~WTH.. ...................

,. .... ,. ......................................................

5 7.0 STRESS RESULTS AND FATIGUE CYCLE LOAPINGS..

....................................

5 8 . 0 PROBABILISTIC FRACTURE MECHANICS EVALXJATION

................................

6 9.0 RESULTS OF ANALYSES/CONCLUSIONS

............................

.. ...............................

7 .... ..........

.............

.....................................................

9.1 Nozzle

Blend Radii

.. .... 8 9.2 Nozzle-to-Shell Weld

.......................................................................................

8 10.0 IaFERENCES

........................

.. .. ,.. .................

8 ....................................................................................

APPENDIX A COMPUTER FILES A-1 List of Tables ... Table 1: Monticello Weld Chemistry

...................................... .A 10 Table 2: Probability of Failure Results Summary .................................................................

10 . File vo, : 1101463, 301 Page 2 of 10 Revision:

0

1.0 OBJECTIVE

This evaluation is to justify the reduction of in-service inspection of the nozzle-to-shell-weld and the nozzle blend radii in the recirculation nozzle (N2) at Monticello Nuclear Generating Plant per code case. N-702 for the extended period of operation.

The N-702 code case, with appropriate technical justification, may be used as an alternative to the requirements of ASME Section XI, Examination Category B-D. 2.0 ' METHODOLOGY The approach was based on the methodology presented in Reference

1. A Monte Carlo simulation was . performed using a variance of the program, VIPER [2] with some modifications as described in the following sections. The VIPER program was developed as part of the program contained in Reference 1 . for the BWR reactor pressure vessel (RPV) shell weld inspection recommendations.

The software was modified into a separate edition, identified as VIPERNOZ, for use in this evaluation. , . The detailed description of the methodology incorporated in the VIPERNIPERNZ program is documented in References 1 and 1 1. 3.0 DESIGN INPUT This analysis is intended for evaluating the reduction of inspection based on the probability of failure in the nozzle-to-shell-weld and nozzle blend radius in N2 nozzles at Monticello Nuclear Generating Plant, Some of the input (e.g., pressure through-wall stress distribution, thermal through-wall stress distribution, and weld residual stress through-wall stress distribution)

Is based on the prior analyses on BWR fleet per References 3 and 4. Others were flom Monticello plant specific described below. Vessel ,Wall Thiclcness

= 5,0625" [5] Vessel Radius = 103.188" [5] Vessel Clad Thickness

= 0.125" [5] Vessel Operating Temperature

= 54g°F [lo, Page 911 Operating P14essure

= 1025 psig [lo, Page 911 Radius to Nozzle-to-shell Weld = 18.25" [6, Figure l],'[17] End of Life Fluence (54 EFPY/GO years) for N2 nozzles = 1.01E18 n/cm2 [7] The weld chemistries are presented in Table 1. For the nozzle blend radius region, since the nozzle is a forging, the number of fabrication flaws was assumed to be 0,l flaws per nozzle, In the weld between the vessel shell and the nozzle, the number of File No.: 1101463,301 Page 3 of 10 Revision:

0 111i Associates, Inc? fabrication flaws was assumed to be 1 per nozzle-to-shell-weld, For both locations, the number of stress corrosion initiated flaws was assumed to be 3 pel4 nozzle OI+ per weld.

All random variabies,were summarized in Table 2 of Reference 8, Most of the input is obtained from Reference 1, except standard deviation fos %Cu and %Ni for nozzle blend radii. They are 0,0447 and 0.068 for

%Cu and %Ni, respectively and were obtained fromBWRVIP-173

[18], 4.0 ASSUMPTIONS The following assumptions are used in the evaluation

[8]: (1) Fabrication flaw is assumed only due to the weld process in the nozzle-to-shell-weld (2) One stress corrosion initiated flaw and 0,l fabrication flaw per nozzle blend radius (3) One fabrication flaw and one stress corrosion initiated flaw per nozzle/shell weld (4) Flaw size distdbution, PVRUF, is assumed, (5) Residual stress at the nozzle/sheU weld is assumed cosine distribution though the wall thickness with a mean of 8 lcsi at surface, , (6) The standard deviation for surface residual stress is assumed to be 5 Icsi. (7) Average upper shelf fiactnse toughness is 200 lcsidin with a standard deviation of 5 lcsijin Several modifications were made to VIPER in order to include the capability to perform the evaluation for nozzle bend radii. The modifications were: Include fatigue crack growth analysis Option to perform stress conosion oraclc growth andlor fatigue crack growth User defined flaw size distribution User defined probability of detection (POD) cuwes for inspection, User defined event occurrence time User defined distribution for selected random parameters User input number of printout for failed and non-failed vessels, The constant for margin term for upper bound vaIues of adjusted reference temperature.

required by Appendix G to 10 CFR Pat2 50 is a user input. Preservice inspection is eliminated, Initial flaw size to include clad thiolness is a uses option.

Improvement in data structure for an,alysis results, The modified software for this project is identified as VIPERNOZ to distinguish fiom the original VIPER software in Reference 1, File No.:

1101463.301 Page 4 of 10 . Revision:

0

6.0 FATIGUE

CRACIC GROWTH The fatigue data for A533-B-1 and A508-2 in reactor water environment are reported in Reference 12 for weld metal testing at R = 0,2 and 0.7. To produce a fatigue crack growth law and distribution for the VPERNOZ software, the data for R= 0.7 was fitted into a form of Paris Law. The Re 0.7 was chosen for conservatism, The curve fit results of the mean fatigue craclc growth law is presented with the Paris Law show11 as foliows: where a = craclc depth n = cycle . AIC = I(max - A comparison to the ASME Section XI [4] fatigue oraclc growth law in reactor water envisonment was done in Reference 8, it shows a very reasonable comparison where Section XI is more conservative on growth rate at high AIC. Using the rank ordered residual plot, it was shown that a Weibyll distribution was more representative for the data, The Weibul residual plot with the linear curve fit of the data is shown below: where y = ln(ln(l/(l-F))

x = In((da/dn),,h,l/(da/dn)m)

F = cumulative probability distribution 7,O STRESS RESULTS AND FATIGUE CYCLE LOADINGS The stress analyses for the nozzle/shell .weld and the nozzle blend radius for the N2 nozzles were presented in Reference 6, The stress analyses were performed for the load cases of unit pressure, and the relevant thermal transients for the N2 nozzles. The through-wall sections were selected based on the thermal transient results, The azimuth locations evaluated were O6, 90") 180° and 270" of the nozzles. Two through-wall sections were selected, Section C is at the location of the weld between the RPV and nozzle. Section D is at the blend radius location of the nozzle. - P File No.: 1101463.301 Page 5 of 10 Revision: 0

@WmWal iutegritY Associates, bc: The load cases analyzed for the N2 nozzles include:

(1) Unit pressure (2) Unit axial load (3) Unit in-plane moment (4) Unit out-of-plane moment (5) Thermal transients depending on the nozzles as described in the following sections For the selected sections in the N2 nozzles (nozzle blend radius and nozzle-to-vessel shell weld), stresses due to the nozzle axial and moment loads are small compared to the pressure and theimal loadings.

Therefore, these load cases were not used in the evaluation.

The thermal transients for the recirculation inlet nozzle are the heat up and sudden pump start of cold recirculation loop, The pressnre is maintained at 1050 psig for the sudden pump stast transient.

For the thermal transients, only the maximum or minimum through-wall stress profiles that produce the largest stress ranges for thermal fatigue crack gsowth are presented and used in the evaluation, The maximum stress among the four azimuth locations was used.

' In this section, the maximum stress is at the 90" and 270" in the hoop direction for the combination of pressure and thermal stresses, The thermal cycles for recirculation inlet nozzle are the number of heathhutdown cycles (288 cycles Refe~ence 16 for an end of operation time of 60 years), and the number of sudden pump sta1.t of cold recirc~rlation loop cycles (10 cycles per page 12 of Reference 13 for an end of operation time of 60 years). 8,O PROBABILISTIC FRACTURE MECHANICS EVALUATION The probabilistic evaluation was performed for the case of 25% inspection rate for period of extended operation (assume 70% inspection rate for initial.40 years of operation at the nozzle blend radii and actual inspection rate for initial 40 years of opesation at the nozzle-to-shell weld location per Reference 19) and 90% inspection coverage, at 10 years interval for 60 years, for the N2 nozzles. For the nozzle blend radius region, a nozzle blend radius crack model, 1141 was used in the probabilistic fsacture mechanics evaluation for the reliability of the in-service inspection progsam. For this location

' and crack model, the applicable stress is the stress perpendiculas to any path cut along the nozzle longitudinal axis. Therefore, the maximum stress among the four azimuth locations (O", 90") 180" and 270") was selected.

For the nozzle-to-vessel shell weld, either a circumferential or an axial craclc could be initiated due to eithes component fabrication (i,e, considering only welding process) or stress corrosion craclcing, From - File No.: 1101463.301 Page 6 of 10 Revision:

0 Reference 1, it is shown that the probability of failure for a circumferential craclc is much less than an axial craclq due to the difference in the stress (hoop versus axial) and the influence function of the craclc model. It is also shown in the through-wall stsess plots in Reference 3 that the difference between the thesn~al hoop stress and the thermal axial stress is not as much compared to the difference between the pressure hoop stress and the pressure axial stress, Therefore, the probabilistic fracture mechanics evaluation for the nozzle and vessel shell weld would concentrate on the axial crack. For the nozzle-to-vessel shell weld, the following craclc model was used in the evaluation: (a) Axial elliptical craclc model with a crack aspect ratio of all = 0.2 , The inspection POD curve is the user input of Figure 42 of Reference 8, with an inspection interval every 10 years, The craclc size distsibution, PVRUF, is shown in Figure 43 of Reference

8. The calculation of stsess intensity factor is at the deepest point of the craclc. The piobability of failure was obtained due to s low temperature over pressurization (LTOP) event at 88 "F and 1150 psi, [IS]. The probability of the LTOP event is 1x10~~ per year 1151. The analyses were performed using VIPERNOZ, a superset of the program VTPER, [2], with the modifications as described in Section 5. The number of simulations was 1 million. 9.0 RESULTS OF ANALYSESICONCLUSIONS Safety evaluation of proprietary EPRI report, dated December 19,2007 states that performing the PFM analysis only for recisculation inlet nozzle (N2) is acceptable it has been demonstrated that the recirculation inlet nozzle is limiting for all sensitivity cases. This conclusion is applicable to both nozzle-to-shell weld and nozzle blend radii. In addition, increased inside surface fluence on reactor vessel components results in decreases of fiacture toughness, increases of reference temperature and increased susceptibility to SCC and LTOP failures, Based on the test data from the parametric studies in BWRVIP-05

[I], increased fluence results in probabilities of failure orders of magnitude higher than unirradiated cases with similar degign parameters in all percentage IS1 exams, The N2 nozzles are the only components applicable to the N-702 code case that have accumulated signiLicant fluence, and the thermal tsansients introduced to the N2 nozzle are as, or more severe, than the transients experienced by the other applicable nozzles, Based on the limiting fluence and stress cases of the N2 nozzle, the results from this analysis shall bound all MNGP nozzle penetrations to the Reactor Pressure Vessel - - File No.: 1101463,301 Page 7 of 10 Revision: 0

9.1 Nozzle

Blend Radii The reliability evaluation is presented for the two cases of in-service inspection.

The probabilities of failure @OF) are summarized in Table 2, For the fist case, 90% inspection coverage over the 60 years of operation and the second case, 25% inspection rate for period of extended operation of 20 years (assume 70% inspection rate for initial 40 years of operation) at nozzle blend radius. The difference between the total conditional failure probabilities for the two cases is less than 1x1~~ per year due to LTOP event. Therefore this analysis de~nonstrates acceptability of reduced in-service inspection per

' code case N-702 at the nozzle blend radii in the recirc~~lation nozzle (N2) at Monticello Nuclear Generating Plant for the extended period of operation, 9.2 Nozzle-to-Shell Weld 1, BWRVIP Repoiat, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)," Electric Power Research 1nstitute.TR-105697, September 1995. 2, VIPER, Vessel Inspection Program Evaluation for Reliability, Version 1.2 (1/5/98), Structural Integrity Associates, 3, SI Calculation W-EPRI-180-301, "RPV Nozzle Stress Analyses," Revision 0. 4. SI Calculation EPRI-180-303, "Deterministic Crack Growth Calculation for BWR Nozzle-to-Shell-welds and nozzle blend radii region," Revision 0, 5. Document NX8290-13, "General Plan Shows Vessel ID (17-2 or 206 Inches) and Vessel Wall Thiclcness (5-1/5") and Cladding (1/8")," SI File Number 1101463.203.

6. SI Calculation 1000720.301, 'Tinite Element Stress Analysis of Monticello RPV Recirculation Inlet Nozzle," Revison 0, 7. DIT 19181-01 "MNGP ~ecirculation Inlet Nozzle-to-shell-welds VIPER Analysis," QF-0545 (IT-E-MOD-11)

Revision 3, SI File 1101463,201. The reliability evaluation is presented for the two cases of in-service inspection, The probabilities of failure (PoF) are summasized in Table 2, For the first case, 90% inspection coverage over the 60 years of operation and the second case, 25% inspection rate for period of extended operation of 20 years (47% I inspection rate for the first 30 years of operation, 78%

inspection rate for the last 10 years of operation I. I for initial 40 yeais of operation per Reference 19) at nozzle-to-vessel shell weld, The difference between the total conditional failure probabilities for the two cases is less than 1x10~~ per year due to LTQP event. Therefore this analysis demonstrates acceptability of reduced in-service inspection per i 1 File No.: 1101463.301 Page 8 of 10 Revision:

0 code case N-702 at the nozzle-to-shell-weld in the recirculation nozzle (N2) at Monticello Nt~clear Generating Plant for the extended period of operation. ,

8, SI Calculation W-EPRI-180-302, "Evaluation of effect of inspection on the probability of failure for BWR nozzle-to-shell-welds and nozzle blend radii region," Revision O,\ 9. Monticello ART Design Input, SI File Number 1000720,204.

10. Document DBD-B. 1, I, "Design Bases Document for Reactor and Vessel Assembly DBD B. 1 .I," Revision C, SI File Number 1101463,203.

11. BWRVIP Report,' "Technical BEisis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend

~adii;" 1003557, October 2002, , SI File Number BWRVIP-01-308, .+ 12, Bamford, W, H., "Application of corrosion fatigue craclc growth rate data to integrity analyses of nuclear reactor vessels," Journal of Engineering Materials and Technology, Vol. 101, 1979 13, Document B,01,01-06, Revision 14, "Operations Manual Section: Reactor and Vessel Assembly B.01.01-06 Figures," SI File Ntunber 1101463,204, 14. Private Communication, P. M. Besuner (Failure Analysis Associates) to P. C,'Riccardella, "Three Dimensional Stress Intensity Factor Magnification Constant for Radial Feedwater Nozzle Craclts," June 1976. 15. NRC, Final Safety Evaluation of the BWR Vessel and Internals, Project BWRVIP-05 Report, TAC # M93925, July 1998, 16, DIT 19181-02 "MNGP Recisculation Inlet Nozzle-to-shell-welds VIPER Analysis," QF-0545 (J?P-E-MOD-11)

Revision 3, SIFile Number 1101463,201, 17, CB&I Drawing No, 7, Revision 9, "12It0 Nozzle MI<, N2 Ah< 17'-2" LD, x 63'-2"'Ins, Heads Nuclear Reactor," Monticello Document No, NX-8920-90, SI File No, 1000720,201.

18, BWRVIP Report, "Evaluation of Chemistiy Data for BWR Vessel Nozzle Forging Materials," 1014995, May2007, SI FileNumber BWRVIP-01-373, 19, DIT 19181-03 "MNGP Recirculation Inlet Nozzle to Shell Welds VIPER Analysis," QF-0545 (FP-E-MOD-11)

Revision 3, SI File Number 1101463.206. File No,:

1101463.301 Page 9 of 10 Revision:

0 SIrucfural Integrity Associates, Inc: aa Table 1: Monticello Weld Chemistry Note: weld and Blend radii respectively, Table 2: Probability of Failure Results Summary , %Cu and %Ni were obtained fiom Reference

7. Initial RTndt were obtalned from References 8 and 9 Nozzle-to-shell-

%Ni 0.99 0.86 %Cu 0.1 0.18 BWR Plant Monticello N2 Nozzle-to-shell-weld Monticello N2 Blend Radii File No,: 1101463.301 Page 10 of 10 Revision:

0 Initial RTndt(F) -65 [8] 40 [Q] Inner Dia (in) 206.4 206.4 Difference in PoF due to LTOP Events between 25% Inspection Rate and 90% Inspection Rate over 60 Years of Operation 8.38E-7 4.45E-8 Shell ThicknessIPath Length (in) 5.0625 9.4845 Conditional POP for 90% In-Service Inspection for 60 Years of Operation 9.24E-3 1.08E-3 Nozzle Blend Radii Nozzle-to-shell-meld Conditional PoF for 25% In-Service Inspection for period of Extended Operation 5.95E-2 3.75E-3

Integrity Associates, Inc? I RID2190,INP ( VIPERNZ input file for 90% inspection oovesage at nozzle blend radii. I File Name RID2125.W Description VIPERNZ input file fo~ 25% inspection coverage at nozzle blend radii. I RID2190.0UT I VIPERNZ output file for 90% inspection coverage at nozzle blend radii, I RIC2125,R\JP RIC219O.INP RID2125,OUT I RIC2125,OUT I VIPERNZ o~~tput flle for 25% inspection coverage at nozzle-to-shell-weld.

I VIPERNZ input file fos 25% lnspectfon coverage at nozzle-to-shell-weld. . VPERNZ input file for 90% inspection coverage at nozzle-to-shell-weld, VIPERNZ output file for 25% inspection coverage at nozzle blend radii, . File No.: 1101463,301 Page A-2 of A-2 RIC219O.OUT VIPI3RNOZlP3.EX.E Revision:

0 VLPERNZ output file for 90% inspcction coverage at noz'zle-to-shell-weld, VIPERNZ executable program