ML20213D840
| ML20213D840 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 10/02/1981 |
| From: | Rubenstein L Office of Nuclear Reactor Regulation |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-0393, CON-WNP-393 NUDOCS 8110280354 | |
| Download: ML20213D840 (17) | |
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OCT 2 1981
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f1EMORAflDUi1 FOR:
R. L. Tedesco, Assistant Director O
- F,s.k. b q'19875.'gl for Licensing
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Division of Licensino
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L. S. Rubenstein, Assistant Director for Core and Plant Systems Division of Systems Integration
SUBJECT:
DRAFT SER F0P,WflP-2 The Reactor Physics Section of the Core Perfomance Branch has prepared the enclosed draft SER for WPPSS fluclear Project flo. 2 (WitP-2).
We have no open issues at this time.
'OrYginni signed by L. S. Rubenstein L. S. Rubenstein, Assistant Director for Core and Plant Systems Division of Systems Integration w
Enclosure:
As stated cc:
A. Schwencer R. Auluck P. Triplett L. Phillips DISTRIBUTi0ft 1
R. Meyer Central Fi3es CPB r/f
Contact:
W. Brooks W. Breo
$ ronu aia no-soi nncu ono OFFICIAL RECORD COPY j
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OCT 2 1931
~ DRAFT SAFETY EVALUATI0t1 REPORT If4PUT WPPSS fiUCLEAR PROJECT t10. 2 (WriP-2)
REACTOR PilYSICS SECTION 4.3 Nuclear Design The applicant has referenced the General Electric Licensing Topical Report NEDE-20944-P (non-proprietary version NE00-20944) for all portions of Section 4.3 except that pertaining to irradiation of the reactor pressure vessel.
This report, entitled " BUR /4 and BWR/5 fuel Design," has been reviewed and approved for such reference (approval letter dated September 30,1977). We find the use of this report for WNP-2 acceptable.
Vessel Irradiation The neutron fluence at the inside surface of the pressure vessel has been calculated by means of a one-dimensional transport theory (discrete ordinates) code in an infinite cylinder geometry. The gross radial power distribution was used in the distributed source option for the code.
The actual flux values were those at the core height having the axial peak flux when the reactor is operating at full power.
The core, core shroud and down-coner region were modeled along with the pressure vessel. An azimuthal peaking factor is calculated by a two-dimensional transport theory method.
The product of the calculated one-dimensional flux, the azimuthal peaking factor and a conservatism factor of 1.5 is taken to be the desired This flux is multiplied by a factor of 1 x 10' (equivalent flux.
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' to 40 years at a capacity factor of 0.8) tc obtain the value of 18 vessel fluence. This value is 1.4 x 10 reutrons per square 4
G electron-volts.
centimeter for neutrons having er.ergy greater than 10 I
We conclude that the analysis of the pressure vessel fluence is This conclusion is based on the fact that state-of-the-acceptable.
art analysis methods are used and that a large conservatism factor l
is applied to the calculated result.
4 15.0 Transient and Accident Analyses 15.4.1 Continuous Rod Withdrawal During Reactor Startup Discussion The Rod Sequence Control System and Rod Worth Minimizer are each designed to enforce a particular rod withdrawal sequence.
Following that sequence would limit the amount of reactivity that could be inserted in one withdrawal action to an amount that would preclude any violation of fuel thermal limits (the programmed withdrawal sequence constitutes normal operation I
during startup).
The probability of a failure in these systems that would permit the continuous withdrawal of a high worth rod is low.
Nevertheless, the consequences of such an i
event have been calculated.
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. The calculation was performed generically by the vendor, General Electric Company, and is reported in the LaSalle County Station Final Safety Analysis Report (Docket 50-341, Section 15.4.1).
The calculation was performed in two steps - first a detailed analysis, including three-dimensional effects, was performed for a rod worth (1.6 perctnt reactivity change) in the upper range of anticipated worth and then a point kinetics calculation was used to extrapolate the results to rod worths to the expected for out of sequence rods. The calculation is performed at one perce.t of full power because calculations have shown that this power level Transient termination is produces the maximum consequences.
assumed to occur by means of the APRM 15 percent power level scram or the degraded (worst bypass condition) 124 scram.
The withdrawal speed is assumed to be the maximum attainable and rod worths up to 2.5 percent reactivity change were In no case was a peak enthalpy greater than 60 analyzed.
calories per gram encountered. We conclude that the analysis of this event is acceptable because a conservative analysis has been performed, conservative rod worth values are analyzed l
and the consequences show a large margin to the acceptance criterion of 170 calories per gram.
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Evaluation Findings The possibilities for single failure of the reactor control system which could result in uncontrolled withdrawal of control rods under low power startup conditions have been reviewed.
The scope of the review has included investigations of initial conditions and control rod reactivity worths and the course of the resulting transient.
The nethods used to determine the peak fuel rod response and the inputs to that analysis have been examined.
The requirements of General Design Criteria 10, 20, and 25 concerning specified acceptable fuel design limits are assumed to be met for this event when the peak fuel enthalpy generated during the transient is less than 170 calories per gram.
The power transient resulting from this event is very narrow (4200 millisecond full width at half maximum) so that fuel enthalpy is an appropriate measure of fuel duty. The staff therefore concludes that the requirements of General Design Criteria 10, 20, and 25 have been met.
This conclusion is based on the following:
The applicant has met the requirement of GDC 10 that the specified acceptable fuel design limits are not exceeded, GDC 20 that reactively control systems are automatically initiated so that specified acceptable fuel design limits are not exceeded, and GDC 25 that single nalfunctions
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of the reactivity control system will not cause the specified acceptable fuel design limits to be exceeded.
These require-ments have been met by comparing the resulting extreme operating conditions for the fuel (i.e., fuel duty) with the acceptance criterion (peak fuel enthalp3 } to assure that fuel rod failure will be precluded for this event.
The basis for acceptance in the staff review is that the applicant's analyses of the maximum low power condition have been confirmed, that the analytical methods and input
.iata are reasonably conservative and that specified acceptable fuel design limits will not be exceeded.
15.4.2 Rod Withdrawal Error at Power Discussion Above a preset power level (approximately 25 percent of full power) the rod withdrawal sequence is no longer enforced by the Rod Sequence Control System or the Rod
" orth Minimizer.
Instead the core is protected against exceeding fuel damage limits by the Rod Block Monitor.
'a' hen a rod is selected for withdrawal the nearest four strings of local power range monitors are also selected.
The outputs from these monitors serve as inputs to the Rod Block Monitor, the output of which is the average of the detector signals.
This output is, then used as input to a trip circuit which is adjusted to block the rod withdrawal
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before fuel damage limits are exceeded.
An analysis is performed to establish the trip setpoint required to accomplish the proper rod block.
The analysis is performed in a conservative manner by assuming the highest worth rod in a pattern to be continuou withdrawn at its maximum speed.
A rod pattern is selected which tends to maximize the consequences though such a
.m pattern would be prohibited during normal operation.
The core is assumed to be operating at rated conditions The two local detector strings having the highest readings ms ww
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are assumed to be inoperable.
The rod to be withdrawn is assumed to be fully inserted prior to its withdrawal The calculation is performed with the BWR Simulator Code which has been reviewed and approved (acceptance letter dated September 22, 1976).
This three-dimensional code is suitable since the po'.ier rise is slow enough to per it i
m the assumption that the neutron and thermal powers can i
be calculated by time-independent methods.The core was assumed to be xenon-free for this calculation in order to maximize the reactivity controlled by control rods The calculatica consists of a number of " snapshots" of the core power distribution as the control rod is withd rawn.
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results of the calculations of linear heat generati examined and curves are on are drawn rates a function and critical power rati as
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Such curves of withdrawal rod.
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dista the highest linear hare drawn for assembli nce for the power ratto during theat generation es containing
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and lowest critical to obtain the maximum tra The curves i
violate established he t vel for the rod which are used a
ratio limits.
genera tion l not The rate or distances is then more limiting of critical power chosen.
the two maximum travel The calculations are also the local detectorsused to which provide inputs to thobtain th Monitor.
of These obtain the output ofare combined in the ap e Rod Block of rod withdrawal.
the Rod Block Monitpropriate manner to or as inoperable local Appropriate a function detectors assumpticns regarding calculation are made.
The set to the value obtaiare plotted and the Rod B results is of the lock "onitor trip travel.
ned for the
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made, we that conservative incalculational m e is conclude that the anal put assumptions accep table.
ysis of are this event is l
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Evaluation Findings The possibilities for single failures of the reactor control system which could result in uncontrolled withdrawal of control rods beyond pormal limits under power operation conditions have been revf&wed.
The scope of the review has included investigations of possible initial conditions and the range of reactivity insertions, the course of the resulting transient and the instrumentation response to the transient. The methods used to determine the peak fuel rod response, and the initial conditions for that analysis have been examined.
The staff concludes that the requirements of General Design Criteria [0, 20, and 25 have been met.
The applicant has met the requirement of GDC 10 that the specified acceptable fuel design limits are not exceeded for the anticipated transient; of GDC 20 that the reactivity cnntrol system is automatically actuated to prevent exceeding the specified acceptable design limits; and of GDC 25 that singic malfunctions in the reactivity control system will not c.ause specified acceptable fuel design Ifmits to be exceeded.
These requirements have been met by comparing 1
the resulting extreme operating conditions and response of the fuel (i.e., fuel duty) witi the acceptance criteria 4
for fuel damage (boiling transition and one percent plastic j
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. strain in the cladding) to assure that fuel rod failure will be precluded for this event.
The basis for acceptance in the staff review is that the applicant's choice of maximum transients for single error control rod malfunctions has been confirmed, that the analytical methods and input data are reasonably conservative and that specified acceptable fuel design limits will not be exceeded.
15.4.7 Operation With an Improperly Loaded Fuel Assembly Discussion Strict administrative controls in the form of previously approved established procedures and, inspection of the loaded core are followed during loading to prevent operation with an improperly loaded fuel assembly.
Nevertheless, an analysis of the consequences of a loading error has been perfor.med.
Two types of loading error are considered-placement of a fuel assembly in an improper location and misorienting a bundle in its proper location.
The enrichment distribution l
Is syrmetric and the water gap surrounding the fuel assembly is uniform for the WNP-2 lattice.
Thus the only effect of misorientation in the first cycle is on the fuel assembly local power peaking distribution which must l
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be accounted for in the thermal hydraulic analysis. The misoriented bundle is tilted from the vertical and the re-sultant change is less than an 0.05 in the critical power ratio.
Normal operation with a misoriented bundle will not cause violation of any fuel damage limits.
The most limiting misloading error occurs at beginning of i
i cycle when a low-enriched bundle is misloaded into the l
high enrichment location of a four-bundle array surrounding l
a local power range monitoring detector. The reading in i
this monitor will be reduced and it will be assumed that i
the readings in the three mirror image four-bundle arrays (which are not instrumented) are the same.
If the instrumented four-bundle array is placed on limits the three mirror image arrays will exceed limits.
The mis-l
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loading error with the maximum calculated consequences for the linear heat generation rate results in an in-crease of this quantity froa the operating limit'of 13.4 j
kilowatts per foot to 14.92 kilowatts per foot in the unmonitored assemblies.
This is far below the approximately 20 kilowatts per foot required to produce a one percent plastic strain in the clad. The misloading 4
event with the most serious calculated consequences for critical power ratio results in a minimum critical power i
ratio reduction of approximately 10.8 percent in the most i
f af fected bundle.
This amount of reduction would yield a
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MCPR of 1.08 if the initial operating MCPR were'1.20.
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. The description given above applies only to the first cycle of operation.
The fuel misloading event is re-l analyzed for each succeeding cycle as part of the cycle loading design.
Evaluation Findings The applicant has evaluated the consequences of the mis-1 loading of a fuel assenbly and has concluded that the most serious misloading would not result in the violation of fuel thermal limits (linear heat generation rate or minimum critical power ratio) when the reactor is operating at Technical Specification limits on thse quantities. This satisfies the requirements of Standard Review Plan Seciton 15.4.7 which requires that any mislooding that cannot be detected by the instruaentation provided in the core causes no fuel damage when the core is operated in the normal
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i 15.4.9 The Rod Orop Accident l
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Discussion i
l The postulated rod drop accident occurs when a rod which has been stuck in the upper portion of the core becomes dis-I connected from the rod drive, the drive is subsequer.tly withdrawn, and the rod becomes unstuck and falls rapidly onto I
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the drive.
This results in a power excursion which_cculd,
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under certain circumstances, result in local fuel dar.ne.
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i I The consequences of the rod drop the amount of reactivity inserted iaccident d 7
dropping rod and on the initial th nto the core b ermal hydraulic conditioni of the core.
Dependence on red drop speed coefficient,
, Doppler feedback the shape of the scram curve and th is less pronounced.
e scram speer has been performed on a generic bT ent Company and is reported in NED0 105 asis b 27, " Rod Drop Accident Analysis for large Boiling Wate and 2 to that report.
r Reactors," and Supplements following conservative assumptioThe c er the ns:
(1) no thermal-hydraulic feedback is as (2) sumed; the least negative Doppler coeffi i c ent which is anticipated is used; (3) the rod drop speed is assumed t o be that measured for the rod design plus three standa d d (4) r eviations; the scram speed is the Technical S (5) the shape of the scram curve is pecification value; and starts with all rods out of the coreas results in the longest delay befo This configuration is inserted into the core. re significant reactivity In addition, the calculational nod l e
For example, the axial flux sha contains conservatism.
constant throughout the excursionpe is assumed t energy deposition in the hot pellet iThis means th s maximited.
. The enthalpy rise in the hot pellet is plotted as a function of the worth of the dropped rod in Nf00-10527, Supplement 1.
l For the design calculation described above a rod worth of I
approximately 1.4 percent reactivity change is required to produce an enthalpy rise of 780 calories per gram, which is our acceptar.:e criterion.
To assess the extent of the conservatism in the assumption of no thermal-hydraulic feedback in the design calculations the staf f consultant, Brookhaven National Laboratory, performed a series of calculations which included this effect.
The results are reported in BNL-NUREG-28109, " Thermal-hydraulic Ef fects on Center Rod Drop Accident's in a Boiling Water Reactor," July, 1980. These results show that if thermal hydraulic feedback is included in the calculations the resulting enthalpy rise is less than 140 calories per gram for a rod worth of 1.4 percent reactivity change. Thus it may be concluded that a large conservatism factor exists in the design calculations.
We have compared the characteristics of the WNP-2 reactor to that used in the generic rod drop accident analysis. The scram reactivity shape for WNP-2 is conservative with respect to the generic study as is the Doppler coefficient
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The rod drop speed and scram speed are the same as or slightly conservative with respect to the values used in the generic analysis. We thus conclude that the generic analysis is applicable to the Susquehanna reactors.
The WNP-2 reactor is provided with a Rod Worth Minimizer and a Rod Sequence Control System to ronitor and enforce the sequence of rod withdrawals in the operational range from cold startup to approximately 25 percent of full power.
In particular, the Banked Position Withdrawal Sequence is enforced.
This sequence has been described in a topical report, NED0-21231, " Banked Position Withdrawal Sequence,"
January, 1977. NED0-21231 contains a generic analysis of potential dropped rod worths and has been reviewed and 3
approved by the staff (approval letter dated January 18, 1978).
In the generic analysis the fuel loading pattern was chosen so as to enhance rod worths compared to what would be expected in a real case.
The values obtained for potential dropped rod worths in the generic analysis were 0.62 percent reactivity change for first cycle rods in the first 50 percent withdrawn with smaller values for succeeding cycles.
During withdrawals of the second 50 percent of the rods the I
maximum worth obtained was 0.75 percent reactivity worth during the first cycle increasing to 0.83 percent in the equilibrium cycle. For comparison the maximum worth calculated for withdrawal of the second 50 percent of rods in the first cycle of WNP-2 was 0.47 percent reactivity chance.
l Using the 0.83 percent value the peak enthalpy from NE001, the design ca The calculacion including the- 0527 is 135 c a m.
predict less than 75 calories rmal-hydraulic feed per gram.
enthalples is expected to produce Neither of thesepeak significant pressure pulse.
claddir.g failure nor a evaluating enviroraental consequeNevert fuel rods suffer cladding failnces, it is assumed that 770 ure.
Eval.uation Findings The staff nas evaluated the ap li p
assumed control rod drop accident cant's analysis calculational techniques, and con and finds the assu I
the calculations predict fuelsequences acceptable. Since calories per gram prompt fuel ruptenthalpies heat transfer to the coolant f ure with consequent rapid was assumed not to occur.
rom finely dispersed nolten U0 a highly local event with no signifiThe app 2
s temperature or pressure cant change in core further conclude that service lik'c co n and Ill of the AS?E Boiler and pre mit C (as defined in Sectio The staff believes that the ca violated.
sufficient conservatism, both in th '
culations contain in the analytical models to e initial assumptions and integrity will be maintainedensure that primary system
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, The staf f further concludes that the requirements of General Design Criterion 28 that the potential amount and rate of reactivity increase be limited in postJlated accidents to preclude greater than limited local yielding in the pressure boundary and preclude significant impairment of the ability to cool the reactor core has been met.
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(,1931 Docket No. 50-397 MEMORANDUM FOR:
A. Schwencer, Chief Licensing Bra ch flo. 2, DL FROM:
R. Auluck, Project fianager Licensing Eranch flo. 2, DL
SUBJECT:
FORTitCOMIflG MEETIflG WITil WASillflGT0!1 PUCLIC P0 HEP, SUPPLYSYSTEft(WPPSS)
DATE & TIIIE:
Wednesday, October 7,1981 8:30 AM LOCATION:
G. E. Conference Room Bethesda, MD PURPOSE:
Discussion of open items in the Auxiliary System Area PARTICIPANTS:
HRC R. Lobed, J. Ridgely and R. Auluck WPPSS_
R. Nelson and Support Staff oricjml siped in R. Auluck, Project l'anager Licensing Branch No. 2 Divis on of Licensing cc: See next page
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Mr. R. L. Ferguson Managing Director Washington Public Power Supply System P. O. Box 968 3000 George Washington Way Richland, Washington 99352 ccs:
Nicholas Reynolds', Esq.
Debevoise & Liberman 1200 Seventeenth Street, N. W.
Washington, D. C.
20036 Richard Q. Quigley, Esq.
Washington Public Power Supply System P. O. Box 968 Richland, Washington 99352 Nicholas Lewis, Chairman Energy Facility Site Evaluation Council 820 East Fifth Avenue Olympis, Washington 98504 Mr. Albert D. Toth Resident inspector /WPPSS-2 NPS c/o U.S. Nuclear Regulatory Commission P. O. Box 69 Richland, Washington 99352 Roger Nelson, Licensing Manager Washington Public Power Supply System P. O. Box 968 Richland, Washington 99352 Mr. O. K. Earle, Project Licensing Supervisor Burns and Roe, Incorporated 601 Williams Boulevard Richland, Washington 99352
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Docket File (s)
IAE Region I liRC PDR I&E Region II Local PDR I&E Region III Branch Reading File I&E Region IV 1151C I&E Region Y TERA TIC NRC
Participants:
E. G. Case D. G. Eisenhut/R. Purple fl. Hughes T. Novak S. Varga T. Ippolito R. A. Clark J. F. Stolz (ORB!4)
R. Tedesco B. J. Youngblood A. Schwencer F. Miraglia E. Adensam (LB!4)
J. R. Miller bcc:
Applicant G. Lainas Service List D. M. Crutchfield B. T. Russell Branch Licensing Branch Ho'. 2 J. Olshinski R. H. Vollmer Project Manger R. Auluck R. J. Mattson S. H. Hanauer Licensing Assistant M. S.ervice T. Murley.
J. P. Knight h
W. Johnston (AD/ Materials & Qualif. Engr}
D. R. fluller P. S. Check W.'E. Kreger i
L. S. Rubenstein F. Schroeder M. L. Ernst-ACRS (16)
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V. Moore B. Grimes 4
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&l(s OCT 0 51981 MEMORAtlDut1 FOR:
Valeria H. Wilson, Management Analysis Branch, Planning & Program Analysis Staff, flRR Walter P. Haass, Chief, Quality Assurance Oranch, FROM:
Division of Engineering
SUBJECT:
F0IA 81-378 - QA 00CUMEi1TATIOl1 FOR WPPSS In accordance with your September 23, 1981, memo we have searched the QAB records for all documentation relating to the staff's evaluation of WPPSS's QA programs relating to their nuclear projects, Wi1P-2,1/4, and 3/5.
All the documents generated by the QAB relating to bither the design / construction or operations QA programs for WPPSS are contained in the Public Document Room.
Orfginal signed by Walter P.Haass Walter P. Haass, Chief Quality Assurance Branch Division of Engineering l
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Icga tive resul t s f inae will be 'repor t ed' to t his of 1 ice by t el ephone.
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Docket No. 50-397 hp UO9b1 MEMORANDUM FOR:
A. Schwencer, Chief Licen' sing Branch po. 2, DL FROM:
R. Auluck, Project Manager Licensing Branch No. 2, DL
SUBJECT:
FORTHCOMING MEETING WITH WASHINGTON PUBLIC POWER SUPPLY SYSTEM (WPPSS)
DATE & TIME:
Friday, Orteh9, -19&t Neua
'L A gefebts 16 19 81 f
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8:30 AM LOCATION:
G. E. Conference Room Bethesda, ljD PURPOSE:
' Discussion of the Preservice Inspection Program (Agenda. Attached)
PARTICIPANTS:
NRC
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M. Hum, J. Gleim, J. Cook (INEL), R.' Auluck WPPSS R. Nelson and Support Staff R. Auluck, Project Manager Licensing Branch No. 2 Division of Licensing
Attachment:
Agenda cc:
See next page hbb%D 0
397 G
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Mr. R. L. Ferguson Managing Director Washington Public Power Supply System P. O. Box 968 3000 George Washington Way Richland, Washington 99352 ccs: Nicholas Reynolds', Esq.
Debevoise & Liberman 1200 Seventeenth Street, N. W.
Washington, D. C.
20036 Richard Q. Quigley, Esq.
Washington Public Power Supply System P. O. Box 968 Richland, Washington 99352 Nicholas Lewis, Chairman Energy Facility Site Evaluation Council 820 East Fifth Avenue Olympis, Washington 98504 Mr. Albert D. Toth
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Resident Inspector /WPPSS-2 NPS c/o U.S. Nuclear Regulatory Commission P. O. Box 69 Richland, Washington 99352 Roger Nelson, Licensing Manager Washington Public Power Supply System P. O. Box 968 Richland, Washington 99352 Mr. O. K. Earle, Project Licensing Supervisor Burns and Roe, Incorporated 601 Williams Boulevard Richland, Washington 99352 m
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AGENDA FOR WASHINGTON NUCLEAR PROJECT NO. 2 MEETING 16 October $,1981 8:30 am to 3:00 pm
Participants:
NRC - R. Auluck, Project Manager M. R. Hun, Materials Engineering Branch J. Gleim, Materials Enginecirng Branch NRC Consultant - J. F. Cook, EG&G Idaho 1.
Objectives of Review 2.
Overview of Schedule and Status of PSI 3.
Examination Sampling Cri.teria and Exemptions Based on IWB-1220 and IWC-1220.
4.
Reactor Vessel Examination - Procedures, Results, Areas that are Impractical to Examine, and Regulatory Guide 1.150.
5.
Piping System Examination - Procedures,. Extent of Examination Coverage, Prac-tice with One-Side Access, Practice with Limitqjians to Examinations, and Resul ts.
6.
Relief Requests E
7.
Safety Relief Valve Discharge Line 8.
NRC Questions and Discussions O
WNP-2 PRESERVICE INSPECTI0tl PROGRAN MEETING TOPICS 1.
For some Class 1 and 2 welds, the WNP-2 preservice inspection plan utilizes Edi-tions and' Addenda of the Code that are not approved by 10 CFR 50.55(a).
For those cases where examinations have not been performed in accordance with approved ASME Code Editions and Addenda, the application must provide a technical justifi-A cation for the use of alternative examinations.
2.
WN states that remote mechanized ultrasonic examination calibration-checks shall be perfonned at intervals not to exceed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> instead of at intervals not to exceed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> as required by Section XI, I-4230.
Confirm that calibra-tion over the intended 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> calibration interval maintains accuracy and sta-bility as required by I-4230.
3.
WNP-2 states that RHR pump. welds are inaccessible.
Is this access limitation only for ultrasonic volumetric examination, or are other examination. methods -
limited also?
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CL 4.
WNP-2 states that no ultrasonic examination on Class 2 piping welds (circumfer-ential and longitudinal) with. piping walls k, inch or less will be performed. This action is not provided for in the 10 CFR 50.55a approved ASME Code,Section XI, 1974 Edition, Summer 1975 Addenda.
The later 1977 Edition up to and including the Summer 1978 Addenda does allow exempting Class 2 piping with walls k, inch or less; however, portions of approved editions or addenda may be used provided that all related requirements of the respective editions or addenda are met.
The pre-
, service and inservice inspection program plan should be modified wherever later 10 CFR 50.55a approved Code editions and addenda are involved to specifically identify the edition and addenda.
5.
WNP-2 states that no ultrasonic examinations will be perfonned on Class 1 piping welds less than 4-inch nominal pipe size.
10 CFR 50.55a has approved the use of the Section XI,1974 Edition up to and including the Summer 1975 Addenda, IWB-1220(b)(1) which states:
"Under the postulated conditions of loss of coolant from the component during normal reactor operation, the reactor can be shut down and cooled down in an orderly manner assuming makeup is provided by the reactor coolant makeup sys tem only. However, in no instance ~ may the size exemp-tion be more than 3 in. nominal pipe size.".
_2-1; hiiditionally, nonnal makeup systems are those sys.,tems thatdave the capability to
. maintain reactor coolant inventory under the respective conditions of startup, hot
~C s tandby, operation or cooldown, using onsite power. - The-tenn onsite power is that
- ' Nower available to the nuclear plant from onsite without any power being supplied
~ from an offsite power distribution Eystem.
The following infonnation concerning piping size exclusions should be provided:
}(a)
Confirm your calculations for liquid and gaseous fluid piping for which Paragraph IWB-1220(b)(1) exemption is invoked.
(b) List the systems and line sizes that were exempted from preservice exami-nation based on Paragraph IWB-1220(b)(1) of the 1974. Edition including
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. Sumer 1975 Addenda,Section XI based on non5al makeu{ systems using only
-onsite power.
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6.-
The WNP-2 preservice inspection program plan does. not.discussi or define as inspec-table systems, the control rod drive cooling waterlines common charging and dis-~
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charging headers, and the control rod insert /withdrawa'l lines.
Identify _the _ ASME
.Cnde. classification of each of these piping systems and discuss the technical basis for your preservice examination requirements.
Those piping system lines that are exempted by IWB-1220(b) and which are parallel pathed into a common header that is larger in size than allowable by IWB-1220(b) should be identified and the comon header lines shall be included in the PSI examination program.
7.
WNP-2 has included in the PSI Program, Table 5.2-12, Items B.4.1 and B.4.5, "Exami-nation Area" as follows:
" Safe end-to-piping welds, safe-ends in branch piping welds circumferen-tial and longitudinal welds to include base metal for a distance of h T or 1 inch, whichever is smaller."
The 1974 Edition of Section XI including addenda through Sumer 1975 cited as the Code being used by WNP-2 requires that base metal examination area include, at least, one-wall thickness beyond the edge of weld.
Discuss the technical basis for your decision for selecting the examination area.
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8.
In Table 5.2-12, Item B-1.4, it is stated that primary nozzle-to-vessel welds and nozzle inner raduis components will be examined from the outside surface.
Confinn that existing procedures and equipment have been qualified on calibration mockups /
standards that enable fatigue cracks to be reliably detected, identified, and characterized.
i 9.
WNP-2 has invoked examination criteria sizing of 2 inches and. smaller for category B-G-2 and larger than 2 inches for category B-G-1 as allowed by the 1978 Addenda to the 1977 Edition of the ASME Code,Section XI.
Confinn that all related exami-nation requirements of the 1978 Addenda to the 1977 Edition are being met.
10.
WNP-2 states in the PSI Program Plan sampling limitations as.follows:
For Item B6.4, Category B-K-1, " Integrally Welded Supports" and Item 86.5, a.
Category B-K-2, " Support Components" of Table 5.2-12, only valves in Cate-gory B-M-1 will be examined.
It is not the in_ tent' of Category B-K-1 and Category B-K-2 to be limited to the sampling &LCategory B-M-1 of ASME, Code,Section XI.
b.
For Item B5.4, Category B-K-1, " Integrally Welded Supports" and Item B5.5, Category B-K-2, " Support Components" of Table 5.2-12, only pumps in Category B-L-1 will be examined.
It is not the intent of Category B-K-1 and Category B-K-2 to be limited to the sampling of Category B-L-1 of ASME Code,Section XI.
For. Item B5.9, Category B-G-2, " Pressure Retaining Bolting 2 Inches and c.
Smaller in Diameter" and Item B6.2, Category B-G-1, " Pressure Retaining Bolts and Studs Greater than 2 Inch Diameter" of Table 5.2-12 only pumps in Category B-L-1 and valves in Category B-M-1 will be examined.
It is not the intent of Category B-G-2 and Category B-G-1 to be limited to the sampling of Category B-L-1 and Category B-M-1 respectively of the ASME Code,Section XI.
Confirm that all required examinations of the above categories will be performed as part of the PSI Program.
I M QO O Dj5 J UCD 009 C DD CG 0600*20D D Q0JE N T-aN
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DocF.et File (s)
I&E Region 1 NRC PDR I&E Region II Local PDR I&E Region III Branch Reading File I&E Region IV HSIC I&E Region V TERA TIC NRC
Participants:
E. G. Case D. G. Eisenhut/R. Purple N. Hughes i
T. Novak S. Varga T. Ippolito R. A. Clark J. F. Stolz (ORBf4)
R. Tedesco B. J. Youngblood A. Schwencer F. Miraglia E. Adensam (LBf4)
J. R. Miller bec:
Applicant G. Lainas Service List D. M. Crutchfield B. T. Russell Branch Licensing Branch No. 2 J. Olshinski r
R. H. Vollmer V{gdect Manger R. Auluck.
R. J. Mattson -
S. H. Hanauer Licensing Assistant M. Service T. Mu rl ey.
J. P. Knight W. Johnston (AD/ Materials & Qualif. Engr)
D. R. Muller P. S. Check W. E. Kreger L. S. Rubenstein F. Schroeder M. L. Ernst ACRS (16)
Ol&E (3)
OSD (7)
Attorney, OELD J. LeDoux, I&E V. Moore B. Grimes O
e
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