ML20140B498

Provides Status of Unresolved Items Discussed at Mechanical Engineering Branch 810609-10 Meeting W/Util.Proposed FSAR Changes Encl

ML20140B498
Person / Time
Site:	Wolf Creek, Callaway
Issue date:	09/05/1981
From:	Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM
To:	Harold Denton Office of Nuclear Reactor Regulation
References
SLNRC-81-090, SLNRC-81-90, NUDOCS 8109140212
Download: ML20140B498 (36)

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SNUPPS IbI.mM' IS Standardia.d Nucl.or Unit 3 S(( 11Ib8\ $-

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Nicholas A. Petrick s che. ch.rry no.d Rockvi% Merytend 20e50

""'" " i gZ Executive Director September 5, 1981 i SLNRC 81-090 File:0541 SUBJ: M_EB Review Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Docket Nos. STN 50-482, STN 50-483, and STN 50-486

Reference:

NRC Summary of June 9-10, 1981 Meeting to Review Mechanical Engineering, dated July 6,1981

Dear Mr. Denton:

On June 9-10, 1981, the SNUPPS applicants met with the NRC's Mechcnical Engineering Branch. The reference reported the results of the meeting. The purpose of this letter is to provide the status of each meeting agenda item not previously resolved.

1. Agenda items 4 and 5 concerned pipe break analysis. SNUPPS FSAR Revisions 5 and 6 provided major updates to this analysis in FSAR Section 3.6. Although the pipe break analysis is not 100 percent complete, sufficient information is thought to be avail-able for the NRC to draw satisfactory conclusions relative to the protection provided in the SNUPPS design against the hazards of postulated piping failures. This conclusion could be confirmed by the NRC in early 1982 when the analysis will be completed and the final FSAR updates to Section 3.6 are provided.

2. The enclosure to this letter provides FSAR changes that will be included in the next revision to the SNUPPS FSAR. These changes resolve meeting agenda items 16, 24, 25, 28, 34, 35, 37, j

and 38.

3. Agenda item 21 concerned the functional capability of piping.

( One clarification in the responsa to item 21 that was contained in the reference is required. In the third paragraph of the response it should be stated that the criteria that are used assure functional capability. The programs serve to check that the criteria are met.

4. Agenda item 29 concerned leak rate testing the isolation for low pressure systems connected to the reactor coolant pressure boundary (Event V). The SNUPPS report on this matter will be transmitted by a separate letter by September 9, 1981.

5. Agenda item 30 concerned the pump and valve test program. The

, program was submitted to the NRC by SLNRC 81-85, dated September 5, 1981. I f1109140212 810905 PDR ADOCK 05000482 A PDR

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. . 3 .

SLNRC 81-90 Page Two j 6. Agenda item 33 concerned an exception to Branch Technical ,

Position MEB 3-1 that SNUPPS suoqested at f.he June 9-10 meeting.

Subsequent to the meeting, SkiUPPS decided tot to take this j, exception concerning splits in Class 1 branch lines. The FSAR

! accurately reflects SNUPPS position on the BTP, and therefore .

this matter should not be an issue.  !

I 6. NUREG-0737, item II.D.1 concerns testing of the pressurizer

} PORV. A change to the SNUPPS commitment concerning this item is

included on revised page 18.2-24 of the enclosed FSAR changes.

i

! 7. NRC question MEB210.2 was transmitted to the SNUPPS applicants in a letter from the NRC's Mr. Youngblood dated August 25, 1981.

l A response to this question (concerning limiting the area of

postulated pipe breaks) will be provided by separate letter by l September 11, 1981.

4 I

Very truly yours,  ;

'. {

RLS Nicholas . Petr ck i- Enclosure cc .J .K . Bry an UE l

D.F. Schnell UE G.L. Koester KGE I

D.T. McPhee KCPL W.A. Hansen NRC/ Cal r T.E. Vandel NRC/WC I

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ENCLOSURE TO SLNRC 81-90 FSAR CHANGF.S

SNUPPS M 6 # 28 3.9(N).l.4.7 Stress Criteria for Class 1 Components and Component Supports All Class 1 components and supports are designed and analyzed for the denign, normal, upset, and emergency conditions to the rules and requirements of the ASME Code,Section III. The design analysis or test methods and associated stress or load allowable limits that will be used in evaluation of faulted conditions are those that are defined in Appendix F of the ASME Code with supplerentary option outlined below: l

a. The test method given in F-1370(d) is an acceptable method of qualifying components in lieu of satisfying the stress / load limits established for the component analysis.

The reactor vessel support pads are qualified using the test option. The reactor pressure vessel support pads are designed to restrain unidirectional horizontal motion, in addition to supporting the vessel. The design of the supports allows radial growth of the vessel but restrains the vessel from horizontal displacements, since tangential displacement of the vessel is prevented at each vessel nozzle.

To duplicate the loads that act on the pads during faulted i conditions, the tests, which utilized a one-eighth linear scale model, were performed by applying a unidirectional static load to the nozzle pad. The load on the nozzle pad was reacted by a support shoe which was mounted to the test fix-ture.

The above modeling and application of load thus allows the maximum load capacity of the support pads to be accurately established.

The test load, L, was then determined by multiplying the T

maximum collapse load by 64 (ratio of prototype area to model area) and including temperature effects in accordance with the rules of the ASME Code,Section III.

The loads on the reactor vessel support pads, as calculated in the system analysis foi faulted conditions, are limited to the value of .e0 L . The tests performed and the limits established T f for the test load method insure that the experimentally obtained valve for L is accurate and that the support pad design i.s T

adequate for its intended function. I 3.9(N)-32 Re . 7 9p

SNUPPS g6M

b. In the design of component supports, member compres-sive axial loads shall be limited to 0.67 times the critical buckling strength, per F-137c(c), of ASME III. ,

Loading combinations and allowable stresses for ASME Code,Section III, Class 1 components and supports are given.in Tables 3.9(N)-2 and 3.9(N)-3.

The methods of load combination for each operating condition are as follows:

Design: Loads are combined by algebraic sum.

Normal, Upset: These loads are used in the fatigue evaluation in accordance with the methods prescribed in the ASME Code.

Loadsets are defined for each transient, including the OBE, and are combined such that the maximum stress ranges are obtained without regard to the order in which the transients occur. (This is discussed in more detail in Section 3.9.1.4.3.)

Emergency: Loads are combined by algebraic sum.

Faulted: For primary equipment, primary equipment supports, and Class 1 branch lines, LOCA and SSE loads are combined  !

using the square-root-of-the-sum-of-the-squares (SRSS) method on a load component basis (i.e. the LOCA F is combined with x  ;

the SSE Fx by SRSS, the LOCA F is cGmbined with the SSE F by y y {

SRSS, and likewise for F , M, M, and Mg ). The sustained z x y loads, such as weight effects, are combined with the SRSS  ;

result by algebraic sum.

For RCL piping, the deadweight moments were added to the LOCA moments prior to the SRSS combination of the LOCA and SSE loads.

3.9(N).2 DYNAMIC TESTING AND ANALYSIS 3.9(N).2.1 Preoperational Vibration and Dynamic Effects Testing on Piping A preoperational piping vibrational and dynamics effects testing program will be conducted for the reactor coolant loop / supports systems during startup functional testing of the SNUPPS units. The purpose of these tests will be to confirm l that the systems have been adequately designed and supported I

for vibrntion as required by Section III of the ASME Code, l Paragrapn NB-3622.3. The tests will include reactor coolant j pump starts and trip If vibrations are observed which, from f

visual examination, appear to be excessive, either: 1) an

'h%5W*PWM~ .9(N)-33 pey, 7 9/81

. .. STUPPS )Mhh 9 6bU instrumented test program will be conducted and the system reanalyzed to demonstrate that the observed levels do not exceed one half the endurance limit at 108 cycles, defined in the ASME code, 2) the cause of the vibration will be elimi-nated, or 3) the support system will be modified to re. duce the vibrations. Particular attention will be provided at those locations where the vibrations are expected to be the largest for the particular transient being studied as per the criteria of the AGME Code referenced above.

It should be noted that the layout, size, etc., of the reactor coolant loop and surge line piping used on SNUPPS is very similar to that employed in Westinghouse plants now in opera-tion. The operating experience that has been obtained from these plants indicates that the reactor coolant loop and surge line piping are adequately designed and supported to minimize vibration. In addition, vibration levels of the reactor coolant pump, which is the only mechanical component that could cause vibration of the reactor coolant loop and surge line piping, are held to acceptable limits. l 3.9(N)-33a Rev. 7 9/81

,, ,. SNUPPS Thus, excessive vibration of the reactor coolant loop and surge l line piping should not be present. However, as added assurance that excessive piping vibration is not present, the piping ad-jacent to the reactor coolant pump will be subjected to visual observation as discussed above. .

3.9(N).2.2 Srismic Qualification Testing of Safety-Related Mechanical Equipment The operability of Category I mechanical equipment must be demonstrated if the equipment is determined to be active, i.e., mechanical operation is relied on to perform a safety function. The operability of active Class 2 and 3 pumps, active Class 1, 2, or 3 valves, and their respective drives, operators, and vital auxiliary equipment is shown by satis-fying the criteria given in Section 3.9(N).3.2. Other active mechanical equipment is shown operable by either testing, analysis, or a combination of testing and analysis. The operability programs implemented on the other active equipment are similar to the program described in Section 3.9(N).3.2 for pumps and valves. Testing procedures similar to the proce-

_* dures outlined in Section 3.10(N) for electrica' equipment are used to demonstrate operability if the component is mechani-cally or structurally complex such that its response cannot be adequately predicted analytically. Analysis may be used if the equipment is amenable to modeling and dynamic analysis.

Inactive seismic Category I equipment is shown to have struc-tural integrity during all plant conditions in one of the following manners: 1) by ana]ysis satisfying the stress criteria applicable to the particular piece of equipment or 2) by test showing that the equipment retains its structural integrity under the simulated test environment.

A list of seismic Category I equipment and the method of qualification used is provided in Table 3.2-1.

3.9(N).2.3 Dynamic Response Analysis of Reactor Internals Under Operational Flow Transients and Steady State Conditions The vibration characteristics and behavior due to flow-induced excitation are very complex and not readily ascertained by analytical means alone. Reactor components are excited by the flowing coolant which causes oscillatory pressures on the surfaces. The integration of these pressures over the applied area should provide the forcing functions to be used in the dynamic analysis of the structures. In view of the complexity of the geometries and the random character of the pressure oscillations, a closed-form solution of the vibratory problem by integration of the differential equation of motion is not always practical and realistic. The determination of the forcing functions as a direct correlation of pressure oscil-lations cannot be practically performed independent of the 3.9(N)-34 Rev. 7 9/81

SNUPPS

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dynamic characteristics of the structure. The main objective is to establish the characteristics of the forcing functions that essentially determine the response of the structures. By studying the dynamic properties of the structure from previous analytical and experimental work, the characteristics of the forcing function can be deduced. These studies indicate that the most important forcing functions are flow turbulence and pump-related excitation. The relevance of such excitations depends on many factors, such as type and location of com-ponent and flow conditions. The effects of these forcing functions have been studied from tests performed on models and prototype plants as well as component tests (Ref. 6, 7, 8, and 14).

The Indian Point No. 2 plant (Docket No. 50-247) has been established as the prototype for a four-loop plant internals verification program and was fully instrumented and tested during hot functional testing. In addition, the Trojan plant (Docket No. 50-344) instrumentation program and the Sequoyah No. 1 plant (Docket No. 50-327) instrumentation program pro-vides prototype data applicable to SNUPPS (Ref. 6, 8, and 14).

The SNUPPS plants are similar to Indian Point No. 2; the only significant differences are the modifications resulting from the use of 17 x 17 fuel, replacement of the annular thermal shield with neutron shielding pads, and the change to the UHI-style inverted top hat support structure configuration.

These differences are addressed below.

a. 17 x 17 fuel The only structural change in the internals resulting from the design change from the 15 x 15 to the 17 x 17 fuel assembly is the guide tube. The new 17 x 17 guide tubes are stronger and more rigid, hence they are less susceptible to flow-induced vibration. The fuel assembly itself is relatively unchanged in mass and spring rate, and thus no significant deviation of internals vibration is expected from the vibration with the 15 x 15 fuel assemblies.

b. Neutron shielding pads lower internals The primary cause of core barrel excitation is flow turbulence, generated in the downcomer annulus (Ref. 8 ) .

The vibration levels due to core barrel excitation for Trojan and SNUPPS, both having neutron shielding pads, are expected to be similar. The coolant inlet density of SNUPPS is slightly lower than Troj an, . and the flow rate is slightly higher. Scale model tests show that the core barrel vibration varies as velocity is raised to a small power (Ref. 7). The difference in fluid density and flow rate ,

3.9(N)-35 Rev. 7 9/81

i SNUPPS gh29 results in approximately 4 percent higher cora barrel vibration for SNUPPS than for Trojan. However, scale model 4

b e

3.9(N)-35a

SNUPPS b test results (Ref. 7) and results from Trojan (Ref. l

6) show that core barrel vibration of plants with l neutron shielding pads is significantly less than that of plants with thermal shields. This infor-mation and the fact that low core barrel stresses and large safety margins were measured at Indian , Point No. 2 (thermal shield configuration) lead to the g

conclusion that stresses less than or equal to those of Indian Point No. 2 will result on the SN?tPPS internals. l

c. UHI-style inverted top hat upper support config-uration  !

The components of the upper internals are excited by turbulent forces due to axial and crossflows in the upper plenum and by pump-related excitations (Ref. 6 and 8). Sequoyah and SNUPPS have the same basic upper internals configuration; therefore, the general vibr;" ion behavior is not changed. The SNUPPS upper ints .als adequacy has been determined from data from  ;

instrumented plant tests at Sey?oyah No. 1, scale model tests, and numerous operating plants. The results of testing at Sequoyah No. 1 (Ref. 14) showed that the components are excited by flow-induced and i pump-related excitations. Analyses of the data indicate that the instrumented components have ade-quate factors of safety, the random flow-induced responses are adequately predicted by scale models, and that the margins are higher with the core in place than during hot functional testing.

In addition, the SNUPPS upper internals configuration was tested in scale model tests, using the same modeling techniques as for the scale model tests of the UHI configuration. The responses of the SNUPPS upper internals have~been calculated using the Sequoyah No. 1 and scale model information. The results show adequate factors of safety for all components.

The original test and analysis of the four-locp configuration is augmented by References 6, 7, 8, and 14 to cover the effects I of successive hardware modifications.

3.9;N).2.4 Preoperational Flow-Induced Vibration Testing of Reactor Internals Because the SNUPPS reactor internals design configuration is well characterized, as was discussed in Section 3.9(N).2.3, it is not considered necessary to conduct instrumented tests of the SNUPPS hardware. The reconmendations of Regulatory Guide 1.20 are satisfied by conducting the confirmatory pre- and 3.9(N)-36 Rev. 7 9/81

., SNUPPS post-hot functional examination for integrity. This examina-tion will include in excess of 30 features illustrated in Figure 3.9(N)-3 with special emphasis on the following areas.

a. All major load-bearing elements of the reactor inter-nals relied upon to retain the core structure in place.

b. The lateral, vertical, and torsional restraints provided within the vessel.

c. Those locking and bolting devices whose failure could adversely affect the s'_ructural integrity of the in-

ternals.

Rev. 7 3.9(N)-36a 9/81

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SNUPPS glhh

d. These other locations on the reactor internal compo-nents which are similar to those which were examined on the prototype Indian Point No. 2, and on Trojan and Sequoyah No. 1.

e. The inside of the vessel will be inspected before and after the hot functional test, with all the internals removed, to verify that no loose parts or foreign material are in evidence.

A particularly close inspection will be made on the following items or areas, using a 5X or 10X magnifying glass, where  !

dppliCable. l

a. Lower internals

1. Upper barrel to flange girth weld.

2. Upper barrel to lower barrel girth weld.

3. Upper core plate aligning pin. Examine bearing surfaces for shadow marks, burnishing, buffing, or scoring. Inspect welds for integrity.

4. Irradiation specimen guide screw locking devices and dowel pins. Check for lockweld integrity.

5. Baffle assembly locking devices. Check for lockweld integrity.

6. Lower barrel to core support girth weld.

7. Neutron shielding pads screw locking devices and dowel pin lockwelds. Examine the interface sur-faces for evidence of tightness. Check for lockweld integrity.

8. Radial support key welds.

9. Insert screw locking devices. Examine soundness of lockwelds.

10. Core support columns and instrumentation guide tubes. Check the joints for tightness and soundness of the locking devices.

11. Secondary core support assembly weld integrity. l 3.9(N)-37 Rev. 7 9/81

SNUPPS dB su The stresses due to the SSE (vertical and horizontal com-ponents) were combined with the blowdown stresses by the SRSS method in order to obtain the largest principal stress and deflection.

These results indicate that the maximum deflections and stress in the critical structures are below the established allowable limits. For the transverse excitation, it is shown that the upper barrel does not buckle during a hot leg break and that it has an allowable stress distribution during a cold leg break.

Even though control rod insertion is not required for plant shutdown, this analysis shows that most of the guide tubes will deform within the limits established experimentally to ensure control rod insertion. For the guide tubes deflected above the no-loss-of-function limit, it must be assumed that the rods will not drop. However, the core will still shut down due to the negative reactivity insertion in the form of core voiding. Shundown will be aided by the great majority of rods that do drop. Seismic deflections of the guide tubes are generally negligible by comparison with the no loss of func-

.* tion limit.

3.9(N).2.6 Correlations of Reactor Internals Vibration Tests With the Analytical Results As stated in Section 3.9(N).2.3, it is not considered neces-sary to conduct instrumented tests of the SNUPPS reactor vessel internals. Adequacy of these internals are verified by use of the Sequoyah and Trojan results, supported by scale model tests. References 7 and 8 describe predicted vibration behavior based on studies performed prior to the plant tests.

l These studies, which utilize analytical models, scale model test results, component tests, and results of previous plant

, tests, are used to characterize the forcing functions and

}

establish component structural characteristics so that the flow-induced vibratory behavior and response levels for SNUPPS are estimated. These estimates are then compared to values dede-Od from plant test data obtained from the Sequoyah and Trojan internals vibration measurement programs.

3.9(N).3 ASME CODE CLASS 1, 2 AND 3 COMPONENTS, COMPONENT SUPPORTS AND CORE SUPPORT STRUCTURES The ASME Code Class components are constructed in accordance with the ASME Code,Section III.

A detailed discussion of ASME Code Class 1 components is provided in Section 3.9(N).l. For core support structures, design loading conditions are given in Section 4.2.2.3.

Loading conditions are discussed in Section 4.2.2.4.

3.9(N)-44 Rev. 7 9/81

SNUPPS MEB 2/(p Loads and deflections imposed on components due to shock and vibration are determined analytically and experimentally in both scaled models and operating reactors. The cyclic stresses due to these dynamic loads and deflections are com-bined with the stresses imposed by loads from component weights, hydraulic forces, and thermal gradients for the determination of the total stresses of the internals. '

The reactor internals are designed to withstand stresses originating from various operating conditions, including thermal shock of the ECCS following a LOCA, as summarized in Table 3.9(N)-1.

The scope of the stress analysis problem is very large, re-quiring many different techniques and methods, both static and dynamic. The analysis performed depends on the mode of opera-tion under consideration.

Allowable Deflections For normal operating conditions, downward vertical deflection of the lower core support plate is negligible.

For LOCA plus the SSE condition, the deflection criteria of critical internal structures are the limiting values given in Table 3.9(N)-12. The corresponding no-loss-of-function limits are included in Table 3.9(N)-12 for comparison purposes with the allowed criteria.

The criteria for the core drop accident is based upon analyses which have to determine the total downward displacement of the internal structures following a hypothesized core drop result-ing from loss of the normal core barrel supports. The initial clearance between the secondary core support structures and the reactor vessel lcwer head in the hot condition is approx-imately 1/2 inch. An additional displacement of approximately 3/4 inch would occur due to strain of the energy absorbing devices of the secondary core support; thus the total drop distance is about 1-1/4 inches, which is insufficient to permit the tips of the RCCA to come out of the guide thimble in the fuel assemblies.

Specifically, the secondary core support is a device which will never be used, except during a hypothetical accident involving the core support (core barrel, barrel flange, etc.).

There are four supports in each reactor. This structure limits the fall of the core and absorbs much of the energy of the fall wnich otherwise would be imparted to the vessel. The energy of the fall is calculated, assuming a complete and instantaneous failure of the primary core support, and is absorbed during the plastic deformation of the controlled volume of stainless steel, loaded in tension. The maximum deformation of this austenitic stainless piece is limited to approximately 15 percent, after which a positive stop is provided to ensure support.

Rev. 7 3.9(N)-69 9/81

SNUPPS

5. Bogard, W. T. and Esselman, T. C.,

" Combination of Safte Shutdown Earthquake and Loss-of-Coolant Accident Re-Sponses for Faulted Condition Evaluation of Nuclear Power Plants," WCAP-9279, farch, 1978.

6. Bloyd, C. N., Claramitaro, W. and Singleton, N. R.,

" Verification of Neutron Pad and 17 x 17 Guide Tube Designs by Preoperational Tests on the Trojan 1 Power Plant," WCAP-8766 (Proprietary) and WCAP-8780, (Non-Proprietary), May, 1976.

7. Lee, H.,

" Prediction of the Flow-Induced Vibration of Reactor Internals by Scale Model Tests," WCAP-8303-P-A (Proprietary) and WCAP-8317-A (Non-Proprietary), July, 1975.

8. Bloyd, C. N. and Singleton, N. R.,

"UHI Plant Internals Vibration Measurement Program and Pre and Post Hot Func-tional Examinations, "WCAP-8516-P (Proprietary) and WCAP-8517 (Non-Proprietary), March, 1975.

9. Takeuchi, K., et al., "MULTIFLEX - A FORTRAN-IV Computer Program for Analyzing Thermal-Hydraulic-Structure System Dynamics," WCAP-8708-P-A, Volumes I and II (Proprietary) and WCAP-8709-A, Volumes I and II (Non-Proprietary),

September, 1977.

10. Fabic, S., " Computer Program WHAM far Calculation of Pressure, Velocity and Force Transients in Liquid Filled Piping Networks," Kaiser Engineers Report No. 67-49-R, November, 1967.

11. Bohm, G. J. and LaFaille, J. P.,

" Reactor Internals Re-sponse Under a Blowdown Accident," First Intl. Conf. on Structural Mechanics in Reactor Technology, Berlin, September 20-24, 1971.

12. Cooper, F. W. Jr., "17 x 17 Driveline Components Tests Phase IB, II, III D-Loop Drop and Deflection," WCAP-8446 (Proprietary) and WCAP-8449 (Non-Proprietary), December, 1974.

i

13. Kraus, S., " Neutron Shielding Pads," WCAP-7870, June, l 1972.

14. Altman, D. A., et al., " Verification of Upper Head Injection Reactor Vessel Internals by Preoperational Tests on the Sequoyah 1 Power Plant," WCAP-9944 (Pro-prietary) and WCAP-9945 (Non-Proprietary), July 1981.

t l

' 3.9(N)-72 Rev. 7 9/81 l

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TABLE 3.9(N)-2 LOADING COMBINATION 3 FOR ASME CLASS 1 COMPONENTS AND SUPPORTS (EXCLUDING PIPE SUPPORTS) l Condition Classificatio_n Loading Combination Design Design pressure, design temperature, i deadwe.ght, operating Basis Earthquake Normal Normal condition transients, deadweight Upset Upset condition transients, deadweight, operating Basis Earthquake Emergency Emergency condition transients, deadweight Faulted Faulted condition transients, deadweight, Safe Shutdown Earthquake or Safe Shutdown Earth-quake and Pipe Rupture Loads Rev. 7 9/81

SNUPPS TABLE 3.9(N)-3 .

ALLOWABLE STRESSES FOR ASME CODE, SECTION III, CLASS 1 COMPONENTS (a)(c)

Operating Component Condition Supports (d)

Classifi ation Vessels / Tanks Piping Pumps Valves Normal NB-3222 (Level A) NB-3653 (Level A) NB-3222 (Level A) NB-3525 (Level A) NF-3222 NF-3231.1(a)

(Level A)

Upset NB-3223 (Level B) NB-3654 (Level B) NB-3223 (Level B) NB-3525 (Level F,) HT-3223 NF-3231.l(4)

(Level B)

Emergency NB-3224 (Level C) NB-3655 (Level C) NB-3224 (Level C) NB-3526 (Level C) NF-3224 NF-3231.l(b)

(Level C)

Faulted NB-3225 (Level D) NB-3656 (Level D) NB-3225 (Level D) (b) NF-3225 NF-3231.1(c)

(Level D)

(a) A test of the components may be performed in lieu of analysis.

(b) CLASS 1 VALVE FAULTED CONDITION CRITERIA Active Inactive A) Calculate P , from para. a) Calculate P,from para.

NB3545.1 with Internal NB3545.1 with Internal Pressure P, = 1.25P, Pressure P, = 1.50 P, P ,ll.5S, P ,22.4S,or 0.7 S u b) Calculate S , from para. b) Calculate S ,from para.

NB3545.2 with NB3545.2 with Cp = 1.5 C = 1.5 p

P, = 1.25P, P, = 1.50 P, Qt =0 Qat

=0 P = 1.3X value 3f P ed P = 1.3X value of P ed ed ed from er?-aions of 3545.2(b) (1) from equations of NB3545.2(b) (1)

S, a 3 S, Sn 53S, P,, P,, P,, Pb' Et ' p' n'S, as m defined by Section III of the ASME Code (c) Limits identified refer to subsections of the ASME Code,Section III.

(d) Also see Appendix 3A, Regulatory Guides 1.124 and 1.130. Q D

Rev. 7 9/81 $

Ps

SNUPPS pfA ,

i 1

TABLE 3.9(N)-4 DESIGN LOADING COMBINATIONS FOR ASME CODE CLASS 2 AND 3 COMPONENTS AND SUPPORTS I") (EXCLUDING PIPE SUPMRTS)

Loading combination (b,c) Design / Service Level Requirements

1. Design pressure, Design design temperature, deadweight

2. Normal condition pres- Service Level A sure, normal condition metal temperature, dead-weight, noz'le loads

3. Upset condition pressure, Service Level B upset condition metal temperature, deadweight, nozzle loads, Operating Basis Earthquake

4. Emergency condition Service Level C pressure, emergency con-dition metal temperature, deadweight, nozzle loads

5. Faulted condition pres- Service Level D sure, faulted condition metal temperature, dead-weight, nozzle loads, Safe Shutdown Earthquake NOTES:

(a) The responses for each loading combination are combined using the absolute sum method. On a case-by-case basis, algebraic summation may be used when signs are known for final design avaluations.

(b) Temperature is used to determine allowable stress only.

(c) Nozzle loads, pressures, and temperatures are those assoc-iated with the respective plant operating conditions (i.e.,

normal, upset, emergency, and faulted), as noted for the component under consideration.

Rev. 7 9/81 I

SNUPPS h TABLE 3.9(N)-5 STRESS CRITERIA FOR SAFETY-RELATED ASME CLASS 2*

AND CLASS 3 VESSELS l

Design / Service Level Stress Limits **

Design and Service Level A o ,5 1.0S (o , or og) + b 5 1.5S Service Level B o,5 1.1S (o , or og) + ob 5 1.65S Service Level C o,5 1.5S (o ,or og) + b $ 1.80S Service Level D o,5 2.0S (o , or og) + ob I 2*48

• Applies for tanks designed in accordance with the AF' Code,Section III, NC-3300.

• • Stress limits are taken from ASME III, Subsections NC and ND, or, for vessels procured prior to the incorporation of these limits into ASME III, from Code Case 1607.

t Rev. 7 9/81

0 SNUPPS TABLE 3.9(N)-6 STRESS CRITERIA FOR SAFETY-RELATED CLASS 2 VESSELS

• Design / Service Level Stress Limits **

Design and Service Level A P,5 1.0S, Pg 5 1.5S, (P , or Pg) + Pb 5 1.5S, Service Level B P,5 1.1S, Pg 5 1.65S, (P,or Pg) + Pb 5 1.655, Service Level C P,5 1.2S, Pg 5 1.8S, (P,or Pg) - Pb 1.8S m

Service Level D P,5 2.05, Pg 5 3.0S, (P , or Pg) + Pb 5 3.0S,

• Applies for tanks designed in accordance with the ASME Code,Section III, NC-3200

• • Stress limits are from ASME III, Subsection NC.

l I

l l

Rev. 7 9/81 t

I

fsf TABLE 3.9(N)-7 STRESS CRITERIA FOR ASME CODE CLASS 2 AND CLASS 3 INACTIVE PUMPS AND PUMP SUPPORTS l Design / Service Level Stress Limits

• Design and Service o ,5 1.0S (o ,or og) + ob 5 1.5S Level A ,

Service Level B o,5 1.1S (o , or og) + ob 5 .65S Service Level C o,1 1.5S (o ,or og) + ob 1.80S Service Level D o,5 2.0S (o ,or o g) + ob 5 2.4S

• Stress limits are taken from ASME III, Subsections NC and ND, or, for pumps procured prior to the incorporation of these limits into ASME III, from Code Case 1636.

Rev. 7 9/81

.- o SNUPPS f,h TABLE 3.9(N)-8 DESIGN CRITERIA FOR ACTIVE PUMPS AND PUMP SUPPORTS l

Design / Service Level Design Criteria *

• Design, Service Level A o,5 1.0S o, + ob 5 1.55 and Service Level B Service Level C o o, + ob m 5 1.2S 5 1.65S Service Level D o 5 1.2S o, + b 5 1.8S m

• The stress limits specified for active pumps are more restrictive than the ASME III limits to provide assurance that operability will not be impaired for any operating condition.

Rev. 7 9/81

SNUPPS

%h He TABLE 3.9(N)-9 STRESS CRITERIA FOR SAFETY-RELATED ASME CODE CLASS 2 AND CLASS 3 VALVES I

Design / Service Level Stress Limits (a,b,c,d, and f) P max I*)

Design and Service Level A Valve bodies shall conform 1.0 to ASME Code,Section III Service Level B o,5 1.lS 1.0 I"m # "L) * "b 5 1.65S Service Level C o m 5 1.5S 1.2 (o , or og) + ob 5 1.808 Service Level D o ,5 2.0f 1.5 (a, or og) + ob 5 2.4S NOTES:

(a) Valve nozzle (piping load) stress analysis is not required when both of the following conditions are satisfied: 1) the section modulus and area of every plane, normal to the flow, through the region defined as the valve body crotch are at least 110 percent of those for the piping connected (or joined) to the valve body inlet and outlet nozzles; and 2) code allowable stress, S, for valve body material is equal to or greater than the code allowable stress, S, or connected piping material. If the valve body material allowable stress is less than that of the connected piping, the required acceptance criteria ratio shall be 110 per-cent multiplied by the ratio of the pipe allowable stress to the valve allowable stress. If unable to comply with this requirement, an analysis in accordance with the design procedure for Class 1 valves in an acceptable alternate method.

(b) Casting quality it.ctor of 1.0 shall be used.

(c) These stress limits are applicable to the pressure re-taining boundary, and include the effects of loads trans-mitted by the extended structures, when applicable.

Rev. 7 9/81 L

SNUPPS g(

TABLE 3.9(N)-9 (Sheet 2)

(d) Design requirements listed in this table are not appli-cable to valve discs, st:ms, seat rings, or other parts of valves which are contained within the confines of the body and bonnet, or otherwise not part of the pressure boundary.

(e) The maximum pressure resulting from upset, emergency, or faulted conditions shall not exceed the tabulated factors listed under P,,x times the design pressure.

If these pressure limits are met, the stress limits in this table are considered to be satisfied.

(f) Stress limits are taken from ASME III, Subsections NC and ND, or, for valves procured prior to the incorporation of these limits into ASME III, from Code Case 1635.

Rev. 7 9/81

SNUPPS ggst3g i

Y TABLE 3.9(B)-3 DELETED t

SNUPPS gg6#37 Y

TABLE 3.9(B)-4 DELETED 1

l l

l l

. .- SNUPPS g 5f REGULATORY GUIDE 1.123 REVISION 1 DATED 7/77 Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants DISCUSSICN:

Refer to Appendix 3A in each Site Addendum.

REGULATORY GUIDE 1.124 REVISION 1 DATED 1/78 Service Limits and Loading Combinations for Class 1 Linear-Type component Supports DISCUSSION:

According to the NRC implementation guidance for this regulatory guide, it is not applicable to the SNUPPS units. However, the following discussion is provided for information purposes.

For ASME Section III components not supplied with the NSSS, the recommendations of this regulatory guide are met as dis-cussed in Table 3.9(B)-14.

The Vestinghouse position with respect to this regulatory guide is as follows.

a. The Regulatory Guide states in Paragraph B.1(b):

" Allowable service limits for bolted connections are derived from tensile and shear stress limits and their nonlinear interaction; they also change with the size of the holt. For this reason, the increases permitted by NF-3231.1, XVII-2110(a), and F-1370(a) of Section III are not directly applicable to allow-able shear stresses and allowable stresses for bolts and bolted connections," and in Paragraph C.4: "This increase of level A or B service limits does not apply to limits for bolted connections."

As noted above, the increase in bolt allowable stress under emergency and faulted conditions is not permitted.

Westinghouse believes that the present ASME Code rules are adequate for bolted connections. This position is based on the following:

It is recognized after extensive experimental work by several researchers that the interaction curve between the shear and tension stress in bolts is more closely represented by an ellipse and not a line. This has been clearly recognized by the ASME. Code Case 1644-6 specifies stress limits for bolts and represents this tension / shear relationship as a nonlinear interaction equation (incorporated into ASME III, Appendix XVII via the Winter 77 Addenda) and has a built-in I

3A-43 p,y, 7 9/81

,- ,. SNUPPS g I

Y safety factor that ranges between 2 and 3 (depending ca whether the bolt load is predominantly tension or shear) based on the actual strength of the bolt as determined by test (Ref: " Guide to Design Criteria of Bolted and Riveted Joints," Fisher and Struik, copyright 1974, John Wiley and Sons, Page 54)'.

Study of three interaction curves of allowable ten-sion and shear stress based on the ASME Code (emer-gency condition allowables per XVII-?l10 and faulted condition allowables per F-1370) and the ultimate tensile and shear strength of bolts (obtained from experimental work published by E. Chesson, Jr., N. L.

Faustino, and W. H. Munse, "High Strength Bolts Subjected To Tension and Shear," Journal of the Structural Division, Proceedings of the American Society of Civil Engineers, October 1965, Pages 155-180) indicates that there is adequate safety margin between the emergency and faulted condition allowables and failure of the bolts.

During their tests to determine the strength and Dehavior characteristics of single high strength bolts subjected to various combinations of tension and shear (T-S), Chesson, et. al. used a total of 115 bolts to ASTM Specificaions A 325-61T and A 354-Grade BC. The A 325-61T, which is a medium carbon steel, had a yield point of 77,000 psi to 88,000 psi and I

ultimate strength of 105,000 psi to 120,000 psi, depending upon the bolt diameter. The A 354-Grade BC, which is a heat treated carbon steel, had a yield point of 99,000 psig to 109,000 psi and ultimate strength from 115,000 psi to 125,000 psi, again I

depending upon the bolt diameter.

I Figure 3A-2 shows the interaction curves for T-S loads on SA-325 bolts. Curve (1) represents the interaction relation (ellipse) permitted by Code Case 1644 (ASME III, Appendix XVII Winter 77 Addenda)

- for service levels A, B, and design condition. Curve (2)

I represents the interaction curve which considers the Code Case 1644 allowables and the increase permitted by XVII-2110(a) for service level C. Curve (3 )

represents the interaction curve which considers the Code Case 1644 allowables and the increase permitted by F-1370(a) for service level D,. Curve (3) is the upper limit of the allowable stresses.

l The design stress limits represented by curves 1, 2, l and 3 for A 325 bolts are then compared against the ultimate strength of the bolts represented by curve 4, l which is based on Chesson's test results. The area between Curve 3 and Curve 4 is the safety margin between the maximum bolt stress under service level D and minimum ultimate strength of the bolt.

l l 3A-43a

.- ,- SNUPPS

[h V

Factor of safety against failure for A 325 bolts for various T-S ratios is shown in Figure 3A-3. The safety factor varies between a minimum of 1.36 and a maximum of 2.29, depending upon the value of T-S ratio. This is based upon the ultimate strength of the bolts from Chesson's test and the allowables obtained from Code Case 1644 and the increase per-mitted by F-1370(a) for service level D. Figure 3A-3 demonstrates that there exists an adequate factor of safety for the complete range of T-S loadings.

From this study it is observed that:

(1) For the emergency condition, the safety factor (ratio of ultimate strength to allowable stress) varies between a minimum of 1.63 and a maximum of 2.73, depending upon the actual tensile stress / shear stress (T/S) ratio on the bolt.

(2) For the faulted condition, the safety factor varies between a minimum of 1.36 to a maximum

. of 2.29, again depending upon actual T/S ratio

• on the bolt.

It is thus reasonable to allow an increase in these limits for the emergency and faulted conditions.

The Westinghouse design of component supports restricts the use of bolting material to the following apolica-tions:

(a) Westinghouse design uses Folting predominantly i.i tension. Oversized holes are generally provided, and a mechanism other than the bolts is provided to take any shear loads. Shear or shear and tension interaction occur only in isolated loca-tions:

(b) Westinghouse bolts are limited to the following materials: A490, SA-354, SA-325, SA-540.

(c) The diameters used range between 1/2 inches and 3 inches. ,

For the emergency condition, Westinghouse will use allowable bolt stresses specified in Code Case 1644, as increased according to the provisions of XVII-2110(a).

For the faulted condition, tensile loads in the bolts shall be limited to 0.7 Su, but not to exceed in any case 0.9 Sy. The allowables are taken at temperature.

In those few cases where bolts are used in shear or tension and shear, ASME Code Appendix XVII - 2460 Requirements will apply with an increase factor that 3A-43b Rev. 7 9/81

,. SNUPPS is defined in Regulatory Guide 1.124 or in Appendix F-1370, whichever is more restrictive. This provides an adequate margin of safety for the Westinghouse design.

b. In Paragraphs B.5 and C.8 of the Regulatory Guide, Westinghouse takes exception to the requirement faat systems whose safety-related function occurs during emergency or faulted plant conditions, must meet level B limits. The reduction of allowable stresses to no greater than level B limits (which in reality are design limits since design, level A, and level B limits are the same for linear supports) for support structures in those systems with safety-related functions occurring during emergency or faulted plant conditions is overly conservative. The primary concern is that the system remains capable of perform-ing its safety function. For active components, this is accomplished through the operability program, as discussed in Section 3.9N.3.2.

c. Paragraph C.6(a) of the Regulatory Guide appears confusing as to what stress limits may be increased for the emergency condition. Westinghouse will interpret this paragraph as follows: "The stress limits of XVII-2000 of Section III and Regulatory Position 3, increased according to the provisions of XVII-2110(a) cf Section III and Regulatory Position 4, should not be exceeded for component supports designed by the linear elastic analysis method."

d. The method described in Paragraph C.7(b) of the Regulatory Guide is overly conservative and incon-sistent with the stress liriits presented in Appendix F.

Westinghouse will use the provisions of F-1370(d) to determine service level D allowable loads for sup-ports designed by the load rating method.

l REGULATORY GUIDE 1.125 REVISION 1 DATED 10/78 Physical Models for Design and Operation of Hydraulic Structures and Systems for Nuclear Power Plants DISCUSSION:

Refer to Appendix 3A of each Site Addendum.

REGULATORY GUIDE 1.126 REVISION 1 DATED 3/78 An Acceptable Model and Related Statistical Methods for the Analysis of Fuel Densification 3A-43c Rev. 7 l 9/81

. SNUPPS M DISCUSSION:

The fuel densification model used for the SNUPPS units is presented in " Fuel Densification Experimental Results and Model for Reactor Application," WCAP-8218-P-A (Proprietary) and WCAP-8219-A (Non-Proprietary), Mareh, 1975, which have been approved by the NRC.

3A-43d Rev. 7 I 9/81

SNUPPS

[h REGULATORY GUIDE 1.127 REVISION 1 DATED 3/78 Inspection of Water-Control Structures Associated with Nuclear Power Plants DISCUSSION:

Refer to Appendix 3A of each Site Addendum.

REGULATORY GUIDE 1.128 REVISION 1 DATED 10/78 Installation Design and Installation of Large Lead Storage Batteries for Nuclear Power Plants DISCUSSION:

The recommendations of this regulatory guide are met. Refer to Section 8.3.2.2.1.

REGULATORY GUIDE 1.129 REVISION 1 DATED 2/78 Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Nuclear Power Plants DISCUSSION:

The recommendations of this regulatory guide are met. Refer to Section 8.3.2.2.1.

REGULATORY GUIDE 1.130 REVISION 1 DATED 10/78 Service Limits and Loading Combinations for Class 1 Plate-and-Shell-Type Component Supports DISCUSSION:

According to the NRC implementation guidance for this regu-latory guide, it is not applicable to the SNUPPS units.

However, the following discussion is provided for information purposes.

For ASME Section III components not furnished with the NSSS, the Class 1 supports are of the linear type and not the plate and shell type. Therefore, this regulatory guide does not apply.

The Westinghouse position with respect to this regulatory guide is as follows.

a. Paragraph B.1 states that increases are not allowed for bolted connections for emergency and faulted conditions. The Westinghouse position is that it is reasonable to allow an increase in the limits for 3A-44 Rev. 7 9/81

SNUPPS g[h${

bolted connections for these conditions. Further justification concerning this position can be found in Item 1 of the discussion on Regulatory Guide 1.124.

l

b. The method described in Paragraph C.7(b) of 'tl}e )

Regulatory Guide is overly conservative and incon-  !

~

sistent with the stress limits presented in Appendix F.

Westinghouse will use the provisions of F-1370(d) to determine service level D allowable loads for supports designed by the load rating method.

l l

3A-44a '

Rev. 7 l

9/81

• 14,247 7 160 (1 ) ALLOWABLE STRE S. i!NDER SERV tCE LE VE L A, 1 SERvlCE LEVEL B. AND DESIGN CONDillCW pef

- ASt4E CODE CASE 1644.

. . (2) ALLOWABLE STRESS UNDER SERvlCE LE VE L C PER

. .; 140 -

ASME CODE CASE 1644 AND APPEhDi s IIII-216 0 ( a ) .

(3) ALLUaABLE STRESS UNDER SERVICE LE vtl D PER ASHE CODE CASE l(44 AhD A P#E hD ix f -1370 (a).

(ul UtilHATE $TRE NGin PER TEST RESULTS (l* OtA.

120

- SA325 BOLTS)

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100 u ULT IMA TE g

STRENGTH 5., SAFETY MARGIN /,

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Q E 80

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$ /

2 w

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60 -

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t /

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3 MAxiMuu 20 A LLOW A BLE ,

FCR SERVICE LE VEL D Al i l 0

40 60 80 100 120 0 20 SHEARSTRESS(KSI) s Figure je d S

Comparison of Tensile Stress for Bolts

i 14,247-8 e *e l

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v .;

E l.0:0.2 1.0:0.67 0.67:1.0 0.2:1.0 4

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I.0:0.0 I.0:0.42 I.0:I.0 0.42:I.0 0.0:I.0 T-S STRESS RATIO Figure m $ 'i f Factor of Safety Against Failure Under Service Level D as a Function of T-S Ratio

~Bb c A v ab =

SNUPPS .-

, i

)

l initiated with a water-solid inlet (loop seal) at the PORVs and a steam bubble maintained in the pressurizer.

Regarding verification of the block valve functionability, the topic is under discussion between PWR utilities and the'NRC staff. Sh"JPPS plans to abide by the final conclusicn; reached by the PW, 2".'i the imC.

1 1872.5.5 Conclusion The SNUPPS plan to demonstrate the operability of the PORVs and safety valves in the SNUPPS plants satisfies the guidance of item II.D.1 in NUREG-0737.

SNUPPS will provide evidence to show the qualification of the block valves by July 1,1982. SNUPPS plans are to demonstrate qualification by use of the generic report to be submitted by PWR utilities in cooperation with EPRI.

)

!u 18.2-24 Rev. 5 7/81

. _ _ . . _