ML20136C125

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Recommends That Task A-43 Re Sump Design Be Maintained on Open Issue Due to Small Sumps,Small Debris Screens & High Recirculation Velocities.Related Info Encl
ML20136C125
Person / Time
Site: 05000000, Vogtle
Issue date: 03/22/1985
From: Serkiz A
NRC
To: Su N
NRC
Shared Package
ML082840446 List: ... further results
References
FOIA-84-663, REF-GTECI-A-43, REF-GTECI-ES, TASK-A-43, TASK-OR NUDOCS 8601030138
Download: ML20136C125 (33)
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March 22, 1985 NOTE TO: Nelson Su h

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FROM: Al Serki:

Docket No.: 50-424

SUBJECT:

Evaluation of Vogtle - 1 Sump Design (Ref. Request for SER Appendix C Input)

Based on my revie,w of. the Vogtle-1 sump design, which is based on design information currently on file, I recommend that Task A-43 be maintained an open issue.

I have discussed the reasons for this recommendation with you previously. The plant specific design uni qui ness, namely: small.

small debris screens and high

sumps, recirculation velocities are the underlying reasons. Enclosure A details further some of my safety concerns and can be used as a SER input 7f sch'edules become overiding.'These matters have been discussed with Containment Systems Branch staff and they plan also to recommend le.aving this an open item pending receipt of information previously requested and further analysis ( or other information ) supporting the conclusion that Loca-generated debris would not result in sump screen blockage and loss loss of NPSH margin to the CSS and RHR' pumps in the' post *LOCA recirculationTperiBd. This matter.'was

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also discussed'with the Applicant on March ~28,1985.

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Kniel, GIB

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Butler, CSB

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Norian,,GIB J.

Shapaker, CSB C.

Li, CSB Melanie Miller, LB-4 I

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e Prepared for Vogtle-1 Docket No.: 50-424 ENCLOSURE A Task'A-43 Cggiainment Emgtgggev gumo Pgtigtmgecg Following a postulated loss-of-coolant accident, water would be collected in the containment emergency sump for use in the long term recirculation mode, thus maintaing core cooling. This water would also be circulated thr'ough the containment spray cooling

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system for removal of heat and fission products within contain-ment. The principal concern is loss of the ability to draw water from the containment emergency sump under post-LOCA conditions--

this leading to the degradation of, or disabling of,the long-term recirculation saf ety train and impairment of decay heat removal.

Two ma-jor concerns have been postulated: (1) adverse hydraulic conditions in the sump (eg. air ingestion, break flow effects,

, vortex formation, etc.) which can lead to loss of pump Net Positive Suction Head - (NPSH), and (2) severe sump screen block-age resulting from LOCA generated insulation debris, which could also lead to loss of NPSH requirements.

a The evaluation.of.such safety concerns hks.been-carried out,' and the Staff's technical findings have been reported in NUREG-0897-(For Comment) and updated (based on,commehts received and further

'experi ments) in the NUREG-OB97,Rev. 1 Draft (May 1984). The destruction of plant insulation by the LOCA-jet is viewed as ~1 safety concern', which could lead to total screen blockage and possibly loss of NPSH. The evaluation of debris blockage is highly plant' specific due the variability of plant designs, insulation. installed and recirculation flow requirements. Air ingestion has been found-to be a function of water elevation and the sump suction inlet Froude number.

e The -Vogtle - 1 sump (s) design differs significantly from the types of designs tested under USI A-43. This plant design employs'four small shallow sumps to supply the CSS and RHR safety trains. Full scale tests have shown that recirculation velocities in the vicinity of these sumps are high (ie 0.4

.to 0.5 ft/sec) and that considerable swirl exsists. This has resulted in the installation of vortex suppressors to mimimize vortening effects and the potential for air ingestion. The A-43 sump test data would also indicate a need for vortex suppression devices in such a shallow sump design.

In addition these recirculation flow velocities wold readily

. transport insulation debris, particularly if a fiberous insolation is used. Vogtle - 1 u'tilizes NUKON insulation, which is a low density fiberglass insulation material. Tests have shown that LOCA jet forces can severely damage and shred such materistis.and transport the 2iberglass log distances.

Furthermore,- transport of shredded fibrous debris occurs at velocities as low as 0.2 ft/sec. Blockage from trans-ported fibrous debris occurs over the entire debris screen

( not just 50% as the applicant has assumed

),

with the thickness build-up occuring over a period of hours and

' dependant on the volume of debris generated and transported

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Vogtle-1 Enclosure A (cont'd)

The Staff has reviewed the Applicant's submittals and analyses

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and arrives at a different conclusion, namely that NPSH margin could potentialy be' lost because of debris blockage. Therefore it is recommended that this issue be maintained an open item pending receipt of additional i nformation previously requested by the staff;and adequate analysis or other data to substantiate the Applicant's conclusions currently filed.

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Prepared for Vogtle-1 d;; GP Docket No.: 50-424 Revi si on 1, April 2,1935 ENCLOSURE A Task A,-43 Cgetgiament Emetgengy SM02 Pgtigtaggcg Following'a costulated loss-of-coolant accident, water would be collected -in the containment emergency sump f or use in the long term recirculation mode, _ thus maintaing core cooling. This vater would-also be circulated through the containment spray cooling system for removal of_ heat and fission products within contain-ment. The principal concern is. loss of the ability to draw water from the containment emergency sump under post-LOCA conditions--

this leading. to the degr edation of, or disabling of,the long-term

' recirculation safety trai.. and impairment of decay heat removal.

.Two' major concerns have been postulated: (1) adverse hydraulic condit, ions,1,n th.e sump _ (eg. air i ng e s,ti on, break flow effects, vortex formation,.etc.) which can lead to loss of pump Net Positive Suction Hesd (NPSH), and (2) severe sump screen block-age resulting from LOCA-generated insulation debris, which could also lead to l'oss of NPSH requirements.

The, evaluation of. such saf ety concerns.has been carried _ ggt.,and the Staff's technical findings have beenLreported in NUREG-0897 (For Comment) and' updated (based on comments received and further experiments) in the NUREG-0897,Rev. 1 Dradt (May 1984): The-

' destruct' ion' of plant-insulation by the LOCA Jet is v'iawed as a

' safety concern which could lead to total spreen blockage and-possibly loss of NPSH. The evaluation of debris blockage is highly plant' specific due the variabili.ty of plant designs, insulation installed and recirculation flow requirements. Air ~

ingestion has been found to be a function of water elevation-and-the sump-suction inlet Froude number.

The.Vogtle - 1 sump (s) ~besign diff ers significantly from the typea.cf designs tested under USI A-43. This plant design employs four small shallow sumps to supply the CSS and RHR safety trains. Full scale tests have shown that recirculation velocitias in the vicinity of these sumps are high (ie 0. 4 to 0.5 ft/sec)-and that considerable swirl exsists. This has resulted in : the installation of vortex suppressors to minimice

'vortexing effects and the potential for air ingestion. The A-43 sumo test data would also indicste a need for vertex suppress' ion devices in such a shallow sump design.

In addition these recirculation flow velocities wold readily transpo.~t1insul ati on. debri s, particularly if a fibercus insulat.on _ i s used. Vogtle

,1 utilices NUKON insulation, which is.a low density fiberglass insulation material. Tests have shown that LOCA jet f orces can severely damag-and shred

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such materials and transport the fiberglass log discances.

Furthermore, transport of shredded fibrous debris occurs at velocities as low as 0.2 ft/sec. Blockage from trans-ported fibrous debris occurs over the entire debris screen-(-not just 50% as the applicant has assumed ).

with the thickners build-up occuring over a period of hours and

.; dependant on the volume of debris generated and transported 1to_the_ sums.

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Revision 1, April 2,1o85-Vogtle-1 Enclosure A (cont'd)

Based on.information submitted by the applicant, and commented on above, the' Staff has found that certain criterion in RG 1.82, Rev. 0 (ie. the recommended sump a,npiveui

.placity of 0.2 ft/sec) have been exceeded and therefore the applicant's assumption c3 a 50% sump blockage is not'necessarily conservative. Such matters have been discussed with the applicant and the Containment Systems Branch (CSB) itas f ound the adequacy-of the Vogtle-1 sump design an open issue pending receipt and review an analysis' of estimated debris generation, sump blockage, and_a reanalysis of pump NPSH margins. The Generic Issues Branch concurs with this finding and will again assist CSB in evaluating analysus that are submitted by the applicant.

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.x VEGP-FSAR-Q Question 210.25 I,

Provide a clarification of the criteria for ther-elimination of circumferential and longitudinal breaks as provided in paragraphs 3. 6. 2.1.2. 2. A.1 and 3. 6.2.1. 2. 2. A.2 which appear to have axial and circumferential stresses transposed.

Response

The project criteria for the elimination of circumferenti'a1 and longitudinal breaks are consistent with the criteria defined in branch technical position MEB 3-1 of standard review plan 3.6.2.

However, these criteria were misstated in paragraphs 3.6.2.1.2.2.A.1 and 3.6.2.1.2.2.A.2 of the FSAR.

These paragraphs m_ N _ revised to reflect the criteria which were actually app 1

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l 51 Q210.25-1 Amend. 12 12/84 I

Table 210.29-1 BREAK OPENING AREAS for Structural Analyses Break Location Calculated Break Areas Break Opening bsed to Develop 2

Areas (in ) U) 11ydraulic Forcing Function (in )

Reactor Pressure Vessel Inlet Nozzle 71 144

-Reactor Pressure Vessel Outlet Nozzle 32 144

-Steam Generator Inlet Nozzle 50 755DE Steam Generator Outlet Nozzle 115 755DE Reactor Coolant Pump Inlet Nozzle 187 755DE Reactor Coolant Pump Outlet Nozzle 33 594DE Loop Closure Weld 128 755DE

. Steam Generator Elbow (Longitudinal) 616 661 for Mass and Energy Release Analyses (for compartment pressurization) i 2

Break Areas (in )

Reactor Vessel Inlet Nozzle 144DE Reactor Vessel Outlet Nozzle 144DE Steam Generator Inlet Nozzle 306DE Steam Generator Inlet Elbow 763SB

- Steam Generator Outlet Nozzle

'436DE Reactor Coolant Pump Inlet Nozzle 336DE Crossover Leg Closure Weld 336DE Reactor Coole.nt Pump Outlet Nozzle 236DE

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Note DE = Double Ended SB = Split Break b.n or qml Ams are d for yef bpMjmend (t)

Grejer 8149Q:1D/010385 5

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VEGP -FSAR-Q

. Question 210.30 Provide.the results of the evaluation of pipe whip and jet and

' impingement effects on safety-related systems,;. components, structures as indicated in table 3.6.2-2.

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Response

f The analysis of pipe whip and jet impingement effects is.being finalized at this time and is scheduled to be complete by h mL 30. 1s65T At that time, a final summary of break

~1ocations, whip restraints, impacted systems, components, structures, and protective devices will be available.

Table 3.6.2-2 will also be updated at that time.

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Q210.30-1 Amend. 12 12/84

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VEGP-ESAR-Q 7

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Question 210.31 Breaks in nonnuclear high energy piping which are not seismically qualified should be postulated at those locations which produce the greatest effect cn1 an essential component or structure.

Provide assurance that the above position has been satisfied.

Response

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r'high energy piping which is a and supported to wi sc...d -afe shutdow quake ' loadings, breaks and hi;b stress points, as are postulated at terminal defined in paragra '

.1.1.B.

In all other nvuu h

energy lines aks are postulated at terminal ends and each interma'.

-e fitting.

This criteria is stated in paragraph 3

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4 VEGP-FSAR-Q Question 210.33 Pa'agraph 3.7.B.1.3 references possible use of damping values r

higher than those listed in tables 3.7.B.1-1 and 3.7.N.1-1, if justified.

If higher damping values were used, provide justification.

Resoonse No mechanical systems or mechanical systems supports are designed using higher damping values than those provided in table 3.7.B.1-1.

g,gg,,, g,7 g 35 ccpric.te and '

1 To date, no equipment has been designed using higher damping values than those provided in table 3.7.B.1-1.

Hi er damping values, if used in the future, will be identified in the associated equipment qualification data packages.

The equipment qualification data packages will, in such cases, identify the documentation that provides the justification for using higher damping values.

Damping values, higher than those specified in table 3.7.N.1-1,

. Damping Values for Seismic Systems Analysis for Westinghouse Supplied Equipment, were not used in the analysis of Westinghouse supplied equipment.

Establishment of damping values, for mechanical componerts by tests, are described in paragraph 3.7.N.1.3 and in the references of this section.

hos It should bg noted that Georgia Power Company 4-r2 I r. the p re e e ns -

ef"requestMagw!;uclear Regulatory Commission approval to use the damping values specified in Code Case N-411 for piping.

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l Q210.33-1 Amend. 12 12/84 l

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VEGP-FSAR-Q

.l b.

Unheated feedwater, whether from the main feedwater system before the feedwater heaters are placed in service or from the auxiliary feedwater system, will be supplied to the steam generator through the auxiliary nozzl'.

This e

reduces the already low likelihood of water hammer occurring in the main feedwater piping and the steam generator main nozzle and

-feedring.

Temperature sensors are provided on the piping c.

close to the steam generator nozzle which will alert the operator of backleakage so that corrective action can be taken.

B.

Auxiliary Feedwater System 1.

Features that prevent draining of the no le region piping into the steam generator:

a.,An upwardly-inclined pipe section is connected to the nozzle inside the steam generator.

b'.

The length of the hori ental flow path at the no :le elevation is minimited by welding a downward directed elbow close to the nostle.

2.

Features that prevent or monitor for backleakage Thru.

r+%e check valves in series are provided g

a.

between the auxiliary no le and the auxiliary feedwater system pump recirculation lines which will restrict backleakage frc.. the auxiliary no::le into the auxiliary feeduater system.

b.

Forward flow will be maintained thrcugh the nozzle as much as possible even during the operation of heatup, hot standby, and cooldown.

Temperature sensors are provided on the piping c.

close to the no :le which will alert the operator of backleakage ss that corrective action can be taken.

"In addition the effects of water hammer in the main feedwater (MFW) and steam hammer in the main steam lines (MSL) were analyzed as follows:

A.

The rapid closure of the MFW check valve and the MSL stop valve is the worst case for water hammer and steam hammer, respectively.

Q210.35-2 Amend. 12

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VEGP-FSAR-Q guestion 210.37 In table 3.7.N.1-1, a damping value of 4 percent for large diameter piping (12 in. NPS and greater) is indicated for the faulted condition and is not consistent with the staff position.

Regulatory Guide 1.61 recommends a damping value of 2 percent for 12-in. piping and 3 percent.for piping greater than 12-in. diameter for the. safe shutdown earthquake.

Furthermore, the staff evaluation'of the Westinghouse Topical Report WCAP-7921 AR as provided in a letter from D.

B. Vassallo (NRC) to R.

Salvatori (W) dated May 16, 1974, specifically stated that the higher damping value was approved for Westinghouse primary coolant loop system ccmponents and large' piping with a configuration similar to that used in the test.

Provide a list ofopiping systems for which the higher damping value was used and additional just.fication for your position.

Response

X.ke 4-percent ~ damping was used in the faulted condition analysis of the7eactor coolant loop piping, and for 12-in. or larger diameter ClasIT pipi.ng attached to the reactor.. coolant loop.

The prer ;rizer surge l'ine~is considere_d as'p' art of the reactor coolant system.

Justification,fo M hE deviation from the guidelines provided in Regulatory Guide l'61 % provided in WCAP-7921-AR,' DanphValues of Nuclear Plant Corsponents.

The evaluation,of-cther large diameter piping ut111:ed the 2-percent and 3-erdent damping for upset and faulted conditions,,

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Q210.37-1 Amend. 12 12/84 9

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a VEGP-FSAR-Q' Q estion 210.39 Provide the procedures used for the design of piping anchors whi'ch separate seismically designed piping and nonseismic Category I piping.

Include a discussion of the loads and load combinations used and how the local pipe wall stresses are considered.

, Response In,the case where an anchor is used to separate Seismic Category I piping systems from piping systems where seismic qualification is not required, the anchor is designed to meet Seismic Category I requirements.

This is in agreement with Regulatory Guide 1.29, paragraph C.3 which states, " Seismic Category I design requirements should extend to the first seismic restraint beyond the defined boundaries.

Those portions of structures, systems, er components that form interfaces between Seismic Category I and nenseismic Category I features should be designed to Seismic Category I requirements."

Loading conditions and load combinations for qualification of piping ccmponents and supports are specified in table 3.9.B.3-1.

In the case of an anchor, the piping analysis for the piping on each side of the anchor is performed independently us;ng the appropriate loading conditions.

Anchor loads are generated for both the upstream and downstream piping runs.

Ancher loads from the nonseismic Category I side include seismic loads due to SSE.

The loads from the two piping runs are then combined and used for the anchor design.

The combination of loads from both sides is normally done c;n::r : tic:lj"Ly an

~

absolute addition of the piping loads.

In the case of a hignly loaded anchor, the appropriate signs of any static loads may be considered,t: clic.inct; :::::::i, c con :rvati;.."' In no case,is a signed static load used to reduce a dynamic load.

4 Local pipe wall stresses are evaluated in accordance with ASME Section III, subsection NB, NC, UD, or ANSI B31.1 as appropriate.

The applicable subsection is determined by the p;pe class.

e t

Q210.39-1 Amend. 12 12/84

,rr1-e-r-

y

4 VEGP-F3AR-Q Question 210.44 Provide a discussion of the design considerations used for safety and relief valve loads and piping reactions.

Include in the discussion the basis for assuring that the valve end loads are acceptable, the support arrangement for the affected piping, and the methodology used to calculate the hydraulic transient forces in the piping due to valve blowdown.

Response

The design considerations used for safety' and relief valve loads and piping reactions are discussed bel'ow.

Calculated valve $$ Lloads are compared to allowable A.

loads identified in the valve specifications and/or in the vendor documentation.

B.

Support arrangements for the affected piping are showr.

in the piping stress calculation folders and on the piping isometrics.

C.

1.

The main steam safety relief discharge piping is designed so that the thrust 1 rce is transferred to the support structure, thus eliminating concerns regarding force transfer to the piping system.

Minor transient forces may exist and were conservatively estimated and fa-tored into the piping system design.

The pressurizer safety relief valveis discussed Scid in revised. d"fMM 2.

$ 9. N. J. 9 /

3.

Other safety relief valves are for thermal relief and the forces associated with them are not significant.

s 4

Q210.44-1 Amend. 12 12/84

pq raph 31#33i VEGP-FSAR-Q JT Pressurizer Safety and Relief System - General Description Special considerations for pressurizer safety and relief valve systems are discussed here.

The pressurizer safety and relief valve discharge piping systems provide overpressure protection for the reactor coolant system.

The three springloaded safety valves, located on top of the pressurizer, are designed to prevent system pressure from exceeding design pressure by more than 10 percent.

The two power-operated relief valves, also located on top of the pressurizer, are designed to prevent system pressure from exceeding the normal operating pressure by more than 100 psi.

A water seal is maintained upstream of each valve to minimize leakage.

Condensate accumulation on the inlet side of each valve prevents any leakage of hydrogen gas or steam through the valves.

The pressurizar safety valves, manufactured by Crosby, are self-actuated, spring-leaded valves with back pressure compensation.

The power-operated relief valves, manufactured by Garrett, are solenoid-operated valves, capable of automatic operation via high-pressure signal or remote manual operation.

The safety valves and relief valves are located in the pressurizer cubicle and are supported by the attached piping which, in turn, is supported by a system of beams, struts, and snubbers.

If the pressure exceeds the setpoint and the valves open, the water slug from the loop seal discharges.

The water slug, driven by high system pressure, generates transient thrust forces.

The valve discharge conditions conservatively considered in the analysis of the pressurizer safety and relief valve piping systems are as follows:

1.

The three. safety valves are assumed to open simultaneously while the relief vallves remain closed.

'r a

2.

The two relief valves open simultaneously while the safety valves are closed.

In addition to these two cases which consider water

~ seal discharge (water slug) followed by steam, solid water from the pressurizer (cold overpressure) is also analyzed.

Q210.44-2 Amend. 12 12/84

l VEGP-FSAR-Q E.

Plant Hydraulic Model When the pressurizer pressure reaches the set pressure (2500 psia for a safety valve and 2350 psia for a relief valve) and the valve opens, the high pressure steam in the pressurizer forces the water in the water seni loop through the valve and down the piping system to the pressurizer relief tank.

Additionally, the power-operated relief valves are subjected to water discharge transients when used for cold overpressure mitigation.

For the pressurizer safety and relief piping system, analytical hydraulic models are developed to represent the conditions described above.

A Westinghouse proprietary computer program ITCHVALVE is used to perform the transient hydraulic analysis for the system.

This program uses the Method of Characteristics approach to generate fluid parameters as a function of time.

One-dinensional fluid calculations applying both the implicit and explicit characteristic methods are performed.

Using this approach, the piping network is input as a series of single pipes.

The network is generally joined together at one or more places by two-or three-way junctions.

Each of the single pipes hac associated with its friction factors, angles or elevation and flow areas.

Conservation equations are converted into characteristic equations.

The ITCHVALVE computer program incorporates special provisions to allow analysis-of valve opening and closing situations.

Fluid acceleration inside the pipe generates reaction forces on all segments of the line.

Reaction forces resulting from fluid pressure and momentum variations are calculated.

These forces can be expressed in terms of the fluid properties available from the transient hydraulic analysis.

The unbalanced forces are calculated using the momentum balance method.

F.

Valve Thrust Analysis The mathematical model used in the seismic analysis is modified for the valve thrust analysis to represent the

-safety and relief valve discharge.

The time-history hydraulic forces from the aforementioned hydraulic analysis are then applied to the pipin'g system lump mass points.

The dynamic solution for the valve thrust is obtained by using a modified-predictor-corrector-integration technique and normal mede theory.

Q210.44-3 Amend. 12 12/84

VEGP-FSAR-Q

)

The time-history solution is found using program FIXFM3.

The input to this program: consists of natural frequencies, normal modes, and applied forces.

The natural frequencies and norma), modes for the pressurizer safety and relief line dynamic model are determined with the WESTDYN program.

Subsequently the time-histor; displacements of the FIXFM3 program are used as input to the WESTDYN2 program to determine the time-history internal forces and deflections at each end of the piping elements.

For this calculation, the displacements are treated as imposed deflections on the pressurizer safety and

-relief line masses.

The solution is stored on tape for later use in the piping stress evaluation and piping support load evaluation.

The time-history internal forces and displacements of the WESTDYN2 program are used as input to the POSDYN2 program to determine the maximum forces, moments, and displacements that exist at each end of the piping elements and the maximum loads for piping supports.

The results from program POSDYN2 are saved for future use in piping stress analysis and support load evaluation.

The major structural analyses programs utilized in this static and dynamic analyses are described in WCAP-8252.

This was reviewed and approved by the Nuclear Regulatory Commission (NRC letter, April 7,

1981, R.

L.

Tedesco to T. M. Anderson).

G.

Comparison to EPRI Test Results Piping load data ha.'e been generated from the tests conducted by EPRI at the Combustion Engineering test facility.

Pertinent tests simulating dynamic opening of the safety valves for representative commercial upstream environments were carried out.

The resulting downstream piping loadings and responses were measured.'

Upstream environments for particular valve opening cases of importance, which envelop the commercial scenarios, are as follows:

1.

Cold water discharge followed by steam - steam between the pressure source and the loop-seal -

cold loc.p seal between the steam and the valve.

2.

Hot water discharge followed by steam - steam

'between the pressure source and the loop seal - hot loop seal between the steam and the valve.

2210.44-4 Amend. 12 12/84 r-m e

,-r-.

e-

VEGP-FSAR-Q 3.

Steam discharge - steam between the pressure source and the valve.

j

. A discussion of the methodology for,-generating the thermal hydraulic forcing functions and a comparison of analytically determined hydraulic force results to test data was presented in the fellow ng article:

L. C. Smith and K.

S. Howe,'"Conparison of EPRI Safety Valve Test Data.with Analytically Determined Hydraulic Results'," The Inter-national Conference on Structural Mechanics in Reactor Technology, Chicago, Illinois, August 22-28, 1983, Volume F, 2/6, pp. 89-96.

A discussien of the methodology utilized in performing a safety valve discharge structural analysis and a comparison of analytical results to structural test results were presenced in the following article.

L.

C.

Smith and T. M. Adams, " Comparison of Analytically Determined Structural Solutions with EPRI Safety Valve Test Results," 4th National Congress on Pressure Vessel and

- Piping Technology, Portland, Oregon, June 19-24, 1983, PVP-Volume 74, pp. 193-199.

H.

System Evaluation In order to evaluate the pressurizer safety and relief valve piping, appropriate load combinations and acceptance criteria were developed.

The load combinations and acceptance criteria are identical to those recommended by the piping subcommittee of the.PWR.PSARV EPR test to ram and are outlined in tables 44-1 and 4-with a definition of load abbreviations provided in table The structural evaluation-of the Class 1 piping is conducted consistent with the rules outlined

-ggAgd in NB-3650 of the ASME Boiler and Pressure Vessel Code Section III.

The piping configuration and

,gf illustrated in ebd downcomer niet support la nur i n t',s figure 2 0.'44-1)threugh 10.44-The piping Ms

  • 65 between nCvEl-ves and the p essurizer relief U hdC.

_ tank is analyzed to satisfy the requirements of

[g*pCI

'the appropriate equations of the ANSI B31-1

,9 fbg4 Code._ jIhe los co inations defined in tables f', M

.44-1 and

10. 4-are utilized.

All static

~

t4 and n mic cases are considerdd in the support Q210.44-5 Amend. 12 12/84

~

- - = -

w+eer

,,9,...,%,r

.w.

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,,,y,_w--

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VEGP-FSAR-Q design and evaluation.

All pressurizer nozzles, valve flanges, and weld attachments are evaluated per ASME Code rules.

A comparison of calculated valve end loads to the design umbrella values tabulated in the piping design specification is conducted to verify acceptability.

Maximum principal stress maximizes bending stress and maximum torsional stress are compared to umbrella operability values if valve operability is required to be demonstrated.

For loading cases where only structural integrity has to be demonstrated, a comparison of maximum principal stresses is conducted.

The accelerations at the center of gravity of the valve are limited to the values provided in the design specification for dynamic loadings.

In summary, the operability and structural integrity of the as-built system is assured for all applicable loadings and load combinations including all pertinent safety and relief valve discharge. cases.

2210.44-6 Amend. 12 12/34

d VEGF-FSAR-2 TABLE -210.?? l'-

LOAD COMBINATIONS AND ACCEPTANCE CRITERIA FOR FRESSURIZER SAFETY AND RELIEF VALVE PIPING - UPSTREAM OF VALVES Ilant/ System Piping Operating Load Allowable Stress

~ Combination Condition Combination Intensity 1

1 Normal N

1.5 Sm 2

Upset N + OBE + SOT 1.8 S,/1.5 S U

y v

i 3

Emergency N + SOT.

2.25 S /1.8 S E

m y

~4 Faulted N + SSE + SOT-3.0 S

=

m b-l

~

1.

Table f.10.44-3'contains SOT definitions and other load abbreviations.

~

2.

SRSS is to be used for combining dynamic load responses.

Amend. 12 12/84

- -- _ - - -----A--s---,.--_

a m----.--_---,a.

s sa

VEGP-FSAR-Q TABLE 210.'."-2 3--

LOAD COMBINATIONS AND ACCEPTANCE CRITERIA FOR PRESSURIZER SAFETY AND RELIEF VALVE PIPING SEISMICALLY DESIGNED DOWNSTREAM PORTION Plant / System Piping Operating Load Allowable Stress Combination Condition Combination Intensity 1

Normal N

1.0 S3 2

Upset N + SOT 1.2 S p

h 3

Upset N + OBE + SOT l'8 S U

h 4

Emergency N + SOT 1.6 S g

h 5

Faulted N + SSE + SOT 2.4 S y

p, 1.

Table -210.00-T"contains SOT definitions and other load abbreviations.

2.

SRSS is to be used for combinit:q dynamic load responses.

Amend. 12 12/84

VEGP-FSAR-2 TABLE 210."

O'"

DEFINITIONS OF LOAD A3BREVIATIONS N

=. Sustained loads during normal plant operation SOT'

= System operating transient SOT..

=~ Relief valve discharge transient U

SOT-

= Safety valve discharge transient z

SOT

= Max (SOT, SOTE),

r transition flou 7

U OBE

= Operating basis earthquake SSE

= Safe shutdown earthquake S3

= Basic material. allowable stress at maximum (hot) temperature S

= Allowable design stress intensity S.

= Yield strength value t

i' Amend. 12 12/84

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Prelluftle' h : Stear / heter Inte face hot to scale CLASS 1 PIPING LAYOUT -

VOGTLE Georgia Power d

'Mf,"LE" RATINC PLANT SAFETY LINES U T FIGURE 210.44-1" 433 9

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ELECTRIC GENERATING PLANT RELIEF LINES GeorgiaPower h ou,T i Auo unir a FIGURE 210.44-2

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UNIT 1 AND UNIT 2 FIGURE Il0. 44 0 -

433 9

VEGP-ESAR-Q Question 210.49 Provide a schedule for the completion of the program for inservice testing of pumps and valves.

The program should

(

contain any relief requests from ASME Section XI requirements together with the justification for _equesting relief.

Response

Due to the requirements of 10CER 50.55 a (g), the code in effect for inservice testing of pumps and valves is indeterminable at this time since 10 CFR 50.55 a (g) requires the use of approved ASME Eection XI Code in effect 12 months prior to operating license date.

It is anticipated that the IST programs will be completed for submittal to the Nuclear Regulatory Commission approximately 3 months prior to the anticipated operating license date of each unit.

However, these manuals will be submitted to the Nuclear Regulatory Commission no later than 6 months following the date of receipt of the operating license for each VEGP unit. 92 # M _ d j=P2 Mn--n 6

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UNITED STATES

['

p, NUCLEAR REGULATORY COMMISSION 7,

E WASHINGTON, D. C. 20555

%,*****j MAY 16 E84 Docket Nos.: 50-424 and 50-425 MEMORANDUM FOR:

Elinor Adensam, Chief Licensing Branch No. 4 Division of Licensing FROM:

Robert J. Bosnak, Chief Mechanical Engineering Branch Division of Engineering

SUBJECT:

REVIEW 0F THE SEISMIC AND QUALITY GROUP CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS AND COMPLIANCE WITH 10 CFR 50.55a FOR V0GTLE ELECTRIC GENERATING PLANI UNITS 1 AND 2 DOCKET h05. 5b-424 and 50-425 During our review of Sections 3.2 and 5.2 of the Vogtle FSAR, the Mechanical Engineering Branch-has identified a number of items that require resolution with respect to FSAR Sections 3.2 and 5.2.

Satisfactory responses to our inquiries as indicated in the enclosure are required in order that we may complete our review.

N Ro'ber'J.Bosnax,Ch3 Mechanical Engineering Branch Division of Engineerinc

Enclosure:

As stated cc: w/ enclosure R. Vollmer, DE J. Knight, DE D. Eisenhut, DL R. Purple, DL T. Novak, DL L

M. Miller, DL H. Brammer, DE D. Terao, DE R. Kirkwood, DE

Contact:

R. Kirkwood, DE:MEB, x28436

_,_.c

Vogtle Electric Generating ' Plant Units 1 and 2 Docket Nos. 50 424 and 50-425 generator (item 6)page 1, the secondary side of the steam In Table 3.2.2-1, 210.5 should be identified as Cuality Group B.

210.6 In Table 3.2.2-1, page 5, the safety injection pump lube oil coolers which are correctly identified as safety-related are classified Safety Class "0" and the construction code is identified as " manufacturer's standards", NRC staff practice requires that these components be classified Quality Groups B or C, that is, constructed to ASME Section III, Class 2 and 3, depending upon the system application.

Identify the principal code used in the construction of the lube oil coolers and demonstrate that the use of this code provides a level of quality commensurate with the safety function of the lube oil coolers.

210.7 In Table 3.2.2-1, page 7, the shell side (Auxiliary Component Cooling Water) ACCW of tne letdown heat exchanger of the (Chemical and Volume Control System) CVCS is incorrectly classified Quality Group D and Safety Class 4.

To be acceptable, the shell side of this ccmponent should be classified Quality Group C (Safety Class 3) and be within the scope of a Quality Assurance Program that is in conformance with 10CFR50, Appendix B.

210.8 In Table 3.2.2-1, page 7, the shell. side, ACCW of the excess letdown heat exchanger of the CVCS is incorrectly classified Quality Group D and Safety Class 4.

To be acceptable, the shell side of this component should.be classified Quality Group C (Safety Class 3) and be within the scope of a Quality Assurance Program that is in confor ance with 10CFR50, Appendix B.

210.9 In Table 3.2.2-1, page 8, the shell side, ACCW of the seal water heat exchanger of the CVCS is incorrectly classified Quality Group D and Safety Class 4.

To be acceptable, the shell side of this component should be classified Quality Group C (Safety Class 3) and be within the scope of a Quality Assurance Program th.It is in confor ance with 10CFR50, Appendix B.

210.10 In Table 3.2.2-1, page 11, the centrifugal charging pump lube oil coolers which are correctly identified as safety-related are classified Safety Class "0" and the construction code is identified as " manufacturer's standards." NRC staff practice requires that these ancillary components be classified quality Groups B or C, that is, constructed to ASME Section III, Class 2 or 3, depending upon the system application.

,:p-

\\

i Identify the principal code used in the construction of the lube oil coolers and demonstrate that the use of this code provides a level of quality commensurate with the safety function of the lube oil coolers.

210.11 In Table 3.2.2-1, page 15, the (Nuclear Service Cooling Water) NSCW pump motor coolers which are correctly identified as safety-related are classified Safety Class "0" and the construction code is identified as " manufacturer's standards". NRC staff practice requires that these -

ancillary components be classified Quality Group C, that is, constructed to ASME Section III, Class 3.

Identify the principal code used in the construction of the NSCW pump motor coolers and demonstrate that the use of this code provides a level of quality commensurate with the safety function of the motor coolers.

210.12 In Table 3.2.2-1, page 16, the following components of the Auxiliary Component Cooling Water System (ACCWS) are incorrectly classified Quality Group D, Safety Class 4, Codes and Standards Designator 5, and Q-List N:

1) ACCWS surge tanks
2) ACCWS pumps
3) interconnecting piping and valves between the above components, the ACCWS heat exchangers, letdown heat exchanger, excess letdown heat exchanger, reactor coolant pumps auxiliary components, and seal water heat exchanger.

To be acceptable the above components should be classified Quality Group C, Safety Class 3, Codes and Standards Designator 3,and Q-List Y.

210.13 In Table 3.2.2-1, page 30, the AFW turbine lube oil coolers which are correctly identified as safety-related are classified as Safety Class "0" and the construction code is identified as " manufacturer's standards." NRC Staff practice requires that these ancillary components be classified Quality Group,C, that is, constructed to ASME Section III, Class 3.

Identify the principal code used in the construction of the lube oil coolers and demonstrate that the use of this code' provides a level of quality commensurate with the safety function of the lube oil coolers.

systems, such as (pages 38, 39, and 40, diesel generator In Table 3.2.2-1, 210.14

1) fuel oil storage and transfer system,

~

(2) jacket coolitig" water system, (3) starting air system, (4) lube oil system, and (5) combustion air intake and exhaust system are classified Quality Group C in SRP's 9.5.4, 9.5.5, 9.5.6, 9.5.7 and'9.5.8.

These SRP's are used by the NRC staff as the basis for determining the

h,gg S

e acceptability of the systems. Therefore, where appropriate, the diesel ger.erator system components in pages 38, 39, and 40 should be revised to reflect the guidance in the SRP's.

210.15 In Table 3.2.2-1, page 42, Post-Accident Sampling System there are three typo errors as follows: Item 4, Seismic Category I, should be 2.

Items 6 and 7, Quality Group B, 4

should be D. Table 3.2.2-1 should be revised accordingly.

210.16 In Table 3.2.2-1, page 52, safety-related piping and valves

'of the instrument and service air system are incorrectly classified Quality Group "NA" and Safety Class "0".

To be acceptable, these components should be classified Quality Group C and Safety Class 3.

210.17 In Table 3.2.2-1, page 58, the hydrogen recombiners are

+

incorrectly classified Quality Group NA.

To be acceptable, these components should be classified Quality Group B.

210.18 In Table 3.2.2-1, page 64, one of the sides of the auxiliary relay room ESF A/C is identified incorrectly as " tube side".

The correct identification is "shell side."

~

210.19 In Table 3.2.2-1, page 87, the electrical penetration assemblies should be classified Quality Group B.

210.20 Verify that all components within the reactor coolant pressure boundary as defined in 10CFR Part 50.2 (V) are 4

classified quality Group A and constructed to Section III, Class 1, of the ASME Boiler and Pressure Vessel Code in compliance with the Codes and Standards Rule, Section 50.55a of 10 CFR Part 50, or as a minimum, are classified Quality Group B and constructed to Section I.II, Class 2, of the code if the components meet the exclusion requirements of the rule.-

210.21 Tables 1.9-1,1.9-2, and 1.9-3 identify certain ASME Code Cases that have been used in the construction of components -

for Vogtle Units 1 and 2.

A number of these Code Cases identified in Regulatory Guides 1.84 and 1.85 are conditionally acceptable to the NRC staff.

Verify that in those instances where conditionally acceptable code cases have been applied in the construction of components you are in compliance with the additional conditions applicable to each conditionally approved Code Case.

210.22 Provide a table in FSAR Section 3.2 of the codes and standards used in the construction of Quality Group A,B, C

~

and 0 components for Vogtle Units 1 and 2.

This tab'le shoulc be similar in format to Standard Review Plan Tabla 3.2.2-1.

1

.. r s:

_4,-'

1 210.23 The hydrogen monitoring subsystem pressure-boundary outside containment is identified in FSAR Section 6.2.5.2.4 as constructed to Quality Group B standards in conformance with R. G.1.26. whereas in Table 3.2.2-1,-- Page 58, and in Figure 9.4.6-2, Sheet 2 of 2, these components are identified as Quality Group C.

in

. Resolve this, consistency in the FSAR.

'210 24 In Table 3.2.2-1 identify the Quality Group, (Safety Class) and the Q-List classification of the reactor coolant pump seals.

e W

4 0

h e

e e

e e

s.

T). :

NOV 141984 Docket Nos:

50-424 h

and 50-425 APPLICANT: Georgia Power Company FACILITY:

Vogtle, Units 1 and 2

SUBJECT:

SUMMARY

OF COMPONENT CLASSIFICATION NEETING HELD SEPTEMBER 19, 1984 On September 19, 1984, the staff met with the applicant and its representatives q

(see Enclosure 1 for participants) to discuss the applicant's classification of y

components and systems at Vogtle.

The discussion in relation to the staff questions a

y follows:

4 Q210.11 This question requests the code used in the construction of the nuclear service cooling water (NSCW) pump motor coolers.

g g

k,hAtthemeeting,theapplicantstatedthattheNSCWpumpmotorcoolersconsist JQg of a ecpper cooling coil with a wall thickness of 0.05 in. The applicant also M _

stated that the cooling coil meets the requirements for seismic Category I piping.

^

Q The staff will evaluate the applicant's complete answer upon receipt of FSAR t

Amendment 10.

h Q 210.12 (including 0210.7 to Q210.9) These questions concern the correct classification of auxiliary component cooling water system (ACCWS) components.

The staff clarified that components which are designed to perform safety functions (RG 1.26) should be classified as Safety Class 3.

The applicant stated that these ACCWS components have been purchased, fabricated and designed to Safety Class 3.

However, they do not plan to install them k

as Safety Class 3.

In order to satisfy the staff, these components must be installed as Safety Class 3 or for those already installed (such as on the lower parts of the Auxiliary Building) the applicant must demonstrate a level of equivalency by examining the installed components to Safety Class.3 criteria.

g The applicant indicated that it would get back to the staff as to whether or not it would be installing the ACCWS components in question to Safety Class 3 and demonstrating a level of equivalency for those already installed.

0210.17 This question deals with classification of the hydrogen recombiners.

The discussion on this question is related to that of Q210.23. The staff stated that systems ar.: components not explicity included in RG 1.26 should not A?pteMr mwsGD FEAR.

gapA)SE Dntpcnav t

N.0?5N

~

[.i

., C riecessarily be labeled as "not applicable" in the appifcant's classification scheme. The applicant's approach does not reflect the guidance in specific

.SRP sections which identify the quality group classification of systems and components which are not included in RG 1.26.

Q210.23 This question regards the classification of the hydrogen monitoring subsystem pressure boundary outside containment.

FSAR paragraph 6.2.5.2.4 indicates that this boundary is constructed to Quality Group B (Safety Class 2) q standards. However, Table 3.2.2-1. Sheet 58 and Figure 9.4.6-2, Sheet 2 3 %, identify this boundary as Quality Group C (Safety-Class 3).

. Q.5*

The staff stated that this difference shoult' be clarified.

The staff further stated h

that this piping should be Quality Group B as in FSAR paragraph 6.2.5.2.4.

The

} N.

applicant plans t,o revise Table 6.2.5-1 to indicate Safety Class 0 with a Footnote W

u to indicate that the boundary is constructed to Safety Class 2 to the outermest isolation valve. The applicant further clarified that the penetration piping N.

and valves are Safety Class 2 and that the remainder of the piping of the hydrogen

$,t monitoring subsystem pressure boundary is Safety Class 3.

The staff indicated

%% that it would get back to the applicant regarding the acceptability of their clarification.

'g Q210.24 This question concerns the classification of the reactor coolant pump seals.

~

The staff stated that the reactor coolant pump seals should be part of the g(

applicant's Q-list. This will ensure that the seals are a part of the operational

.Q quality assurance program with strict controls over such areas as documentation y

and procedures in seal replacement.

The applicant indicated that it would get back to the staff as to whether or not it would be adding the seals to the Q-list.

Melanie A. Miller, Project Manager Licensing Branch No. 4 Division of Licensing

Enclosure:

As stated C

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