ML20236Y302
| ML20236Y302 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, 05000000 |
| Issue date: | 12/02/1987 |
| From: | Ainger K COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML19302D141 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM 3911K, NUDOCS 8712110228 | |
| Download: ML20236Y302 (17) | |
Text
_ _ _ _ _.. _ _
[ N C:mm2nw cith Edis:n t
$ / One First National Plaza, Chicago, Illinois N>
. Address Reply to: Post Office Box 767 Chicago, lilinois 606C 3 0767 December 2, 1987 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Subject:
Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 Pressurizer Relief and Safety Valves NRC Docket Nos. 50-454/455 and 50-456/457 References (a): March 26, 1987 letter from L.N. Olshan to D.L. Farrar (b): March 26, 1987 letter from J.A. Stevens to D.L. Farrar Gentlemen:
~ References (a) and (b) requested additional information concerning NUREG-0737, Item II.D.1, Performance Testing of Relief and Safety Valves.
Attachment A and attachments 1-10 of this letter provide the response to the request for additional information.
Please direct any further questions regarding this matter to this office.
Very truly yours, K. A. Ainger Nuclear Licensing Administrator Attachments cc: ' Byron Resident Inspector Braidwood Resident Inspector NRC Region III Office f
ao 7\\
.l b,,
3; ( b 3911K fh I O y aa 1 0 2 1 g g [rd 4 o P. j
Attachment A Byron /Braidwood (} Plant Specific Responses to NRC Questions Regarding NUREG-0737. II.D.1 1. PORV Block Valve (Byron 1 and 2; Braidwood 1 and 2) The submittal did not identify the PORV block valves used at Byron /Braidwood, Units 1 and 2. The PWR Valve Data Package (Reference 6) listed the PORV block valves used at Byron /Braidwood as Westinghouse 3 in, gate valves, Model 03000GM88-FNH0000SW75 with Limitorque SB-00-15 operators. This model number is slightly different from the Westinghouse gate valves tested by EPRI. Identify the differences between the PORV block valves used at Byron / Braidwood and those tested by EPRI. Discuss what effect these differences may have on valve operability. Discuss the operability of the Byron / Braidwool block valves in both steam and water discharge conditions. The EPRI tests indicated that the Westinghouse block valves experienced closure problems against full steam flow. Discuss why similar closure problems will not occur with the Byron /Braidwood block valves or what steps have been taken to ensure the valves will fully close at Byron /Braidwood, Units 1 and 2. Response (Byron 1 and 2; Braidwood 1 and 2) "he block valves used at Byron /Braidwood are Westinghouse 3 in. gate valves, Model 03000GM88FNH0J, built to assembly drawing 1154E423. This model was represented in the EPRI/ Marshall Block Valve Tests by Westinghouse Model 03000GM88FNBODO, built to assembly drawing 8374D34. O The differences in the two models are the weld end preparation, external limit switches and motor insulation. The tested model had Class B insulation, while Byron /Braidwood have LR. Additionally, the plant valves have external NAMCO limit switches while the tested valve did not. These differences have no impact on valve operability. Although the same design pressure and temperature apply to both models, because different ASME Boiler and Pressure Vessel Codes were used, the nominal pressure rating and minimum wall requirements are different. This did not, however, result in any physical changes to the hardware and is not relevant to operability. The EPRI tests conducted were on steam only, although the block valve experiences the same range of inlet conditions as the PORV. It is assumed that these full flow steam tests represent the most severe of the expected modes of operation. The Westinghouse valve with the Limitorque SB-00-15 motor did experience closing problems in this test series. The problem was resolved satisfactorily with increased torque switch settings. The PORV block valves at Byron /Braidwood subsequently received changes to their motor operators. The torque switch settings were adjusted to their maximum values, the pinion gear ratios were adjusted, and the wiring changed from torque controlled to travel (limit) controlled stroking. These changes provide the maximum possible assurance that closure problems will not occur (~3 in the plant. l \\-) l i
pegs'2 - Attachment A. Byron /Brnidwood, plant Specific Responses to NRC Ouestions Rngarding NUREG-0737. II.D.1 2. Loop Seal Temperature (Byron 1 and 2; Braidwood 1 and 2) The Licensee stated that the installation of the safety / relief valve piping and supports was not completed at Byron /Braidwood Units 1 and 2 at the time of the submittal in 1982. At the time, the Licensee was considering using hot loop seals at the inlets of the safety valves. No additonal information has been received since the 1982 submittal. Therefore, confirm whether a hot loop seal was used for the Byron /Braidwood safety valve. Because the fluid forces acting on the piping system can be significantly affected by the loop seal temperature assumed in the thermal-hydraulic analysis, provide a comparison of the loop seal temperature profile used in tha thermal hydraulic analysis with the actual temperature measured at the plant. Response (Byron 1 and 2; Braidwood 1 and 2) Hot loop seals have been utilized at Byron Units 1 and 2 and Braidwood Units 1 and 2. The design of the insulation boxes which were utilized is presented in calculation 368-0283-001 (Attachment 10). The calculation applies to the Byron 1 and 2 and Braidwood 1 and 2 loop seals. O v l 1 O r l
Page 3 - Attachmtnt A. Byron /Braidwood. Plant Specific Responses to NRC Questions Rwnerding NUREG-0737. II.D.1 OV. Safety Valve Ring 3 Settings (Byron 1 and 2; Braidwood 1 and 2) The submittal stated that Byron /Braidwood Units 1 and 2, used the manufacturer recommended safety valve ring positions but did not identify the actual ring positions. Give the ring settings for the three safety valves at each unit of the Byron /Braidwood plant. Provide the ring settings as referenced from the level position so that the ring settings at the plant can be compared to those used in the EPRI tests. Response (Byron 1 and 2; Braidwood 1 and 2) The following are the original, as-shipped Byron and Braidwood safety valve ring positions. The nozzle ring position is the same with regard to the level or highest locked position. The guide ring settings are provided both with respect (W/R/T) to level and highest locked positions. While these numbers are not identical to those shown by EPRI as " manufacturer's recommended" ring position, these numbers were obtained as a result of the same dynamic test procedure, to the same acceptance criteria by the manufacturer. Differences in internal part dimensional stackups (within accepted tolerances) require that this testing be done on each valve, and settings vary from valve to valve. Therefore, the ring settings at Byron and Braidwood are valve specific manufacturers ring settings and should provide the same performance as the test valves which also had " manufacturer's recommended" position, even though the numbers vary. () Guide Ring Guide Ring Valve Serial Number Nozzle Ring (W/R/T Locked) (W/R/T Level) Byron Unit 1 N-56964-00-0030 -18 -300 -155 N-56964-00-0031 -18 -250 -107 N-56964-00-0032 -18 -250 -101 Byron Unit 2 N-56964-00-0047 -18 -250 -106 N-56964-00-0048 -18 -250 -100 N-56964-00-0049 -18 -250 - 98 Braidwood Unit 1 N-56964-00-0053 -18 -310 -125 l N-56964-00-0054 -18 -225 - 79 j N-56964-00-0055 -18 -225 - 88 l 1 l Braidwood Unit 2 l N-56964-00-0071 -18 -275 -112 N-56964-00-0072 -18 -250 -119 () N-56964-00-0073 -18 -250 - 96 1 l l lL__________________________________________
Page 4 - Attachm+nt A. Byron /Braidwood. Plant Specific Responses to NRC Ouestions Rinerding NUREG-0737. II.D.1 ,RV 4. Safety Valve Inlet Pressure Drop and Backpressure (Byron 1 and 2; Braidwood 1 and 2) In discussing the periJrmance of the 6M6 safety valve, the Licensee stated that the upstream piping length and shape were essentially the same for the EPRI test and Byron /Braidwood plant configurations (Reference 1). This implies that the pressure drop (and pressure rise) in the Byron /Braidwood safety valve inlet piping should be comparable to those of the EPRI tests. Reference 1 also stated that the EPRI backpressure were significantly higher than the Byron /Braidwood backpressure. In order to make a direct quantitative comparison of the upstream and downstream fluid flow charac' eristics of the safety valve piping, the Licensee should present the followltg items, a. The total pressure drop and pressure rise in the Byron /Braidwocd, Units 1 and 2, inlet piping associated with safety valve opening and closing. Provide a numerical comparison of the Eyron/Braidwood pressure drop and pressure rise values with those calculated for the EPRI test configuration. The total pressure drop should include both the frictional and acoustic wave components evaluated under steam discharge conditions. b. The maximum backpressure for Byron /Braidwood, Units 1 and 2, safety valve discharge piping. Response (Byron 1 and 2; Braidwcod 1 and 2) The total pressure drop and rise for safety valve opening and closing a. is calculated in V-EC-410 ( Attachment 5). This includes both the frictional and acoustic wave components of the pressure change, and is calculated for steam discharge. Only the longest of the twelvo lines at the 4 units was modeled, as it provides the greatest pressure change. The results are: Byron /Braidwood EPRI 6M6 "G" Opening 234.84 psi 263 psi Closing 119.69 psi 181 psi The max
- mum dynamic backpressure calculated was 533 psia for Byron /
Braidwood. The tests on the 6M6 valve at EPRI included backpressure from 227 psia (test 910) to 710 psia (test 929), with an average value of 471 psia over 24 t u ts. The report summary (NP-2770-LD Vol.6 pg.S-5) concluded that the "Cecsby valve steam test blowdowns....were relatively unaffected by typical pWR plant backpressure." The Byron /Braidwood backpressure were well within the range tested, and are expected to have i no affect on valve operability. b. The maximum calculated backpressure for the safety valves was 533 psia. This is noted in Section 6.1.1 of Attachments 1 thru 4.
l 'Pzgo 5 - Attachmtnt A. Byron /Braidwood. Plant Specific Responses to NRC Ouestions j M*Rerding NUREG-0737. II.D.1 O 5. Low Temperature Overpressure Events (Dyron 1 and 2; Braidwood 1-and 2) The submittal stated that the inlet fluid conditions tested by EPRI envelop range of. conditions achiavable at Byron /Braidwood in'a low temporture overpressurization event. It did not provide details on the pressure and temperature conditions for the PORV during plant shutdown operation. Identify the low pressure set point of the PORV and the temperature at which' the. low set point is enabled. Identify the bounding inlet' fluid conditions (pressure and temperature) for low temperature overpressure transients including steam and water discharges. State whether the minimum operating 0 l-pressure limit for the Byron /Braidwood PORV is 300 psig at 100 to 400 F as shown in Figure 5-1 of Reference 7. Identify the test data that demonstrate valve operability over-this range of fluid conditions. Response (Byron 1 and 2; Braidwood 1 and 2) Attached is an excerpt from the Byron and Braidwood " Precautions, Limitations .and Setpoints" document, (Attachment 6) showing the COMS setpoints. To bound the conditions, a maximum rise of 100 psi over the maximum setpoint, and a maximum pressure undershoot of 120 psi below the niinimum setpoint is 0 assumed. This reruits in bo~nding conditions of 2450 psig at 450 F, and 0 350 psig at 70. The PORV used at Byron and Braidwood is the Copes-Vulcan D-100-160 with 17-4PH cage and 316SS stellite clad plug. EPRI test results for this valve are available in NP-2144-LD "EPRI-Marshall Relief Valve Test Report" and NP-2670-LD, Vol. 8 "EPRI/Wylie PORV test Report". 0 The Wylie tests consisted of one steam test at 2715 psia and 682 F; 0 0 6 water tests from 675 psia at 105 F to 2535 psia at 647 F; and 2 transition tests. The Marshall test series were 4 steam tests, at approximately 2430 psig of saturated steam. In all tests, the PORV operated satisfactorily. As most of the potential COMS events occur before a steam bubble is drawn in the pressurizer, the water tests are most applicable to the COMS conditions. j The 6 Wylie water tests, 72-CV-316-3W, 73-CV-316-4W, 74.-CV-316-5W, j 76-CV-316-6W. 76-CV-316-2W 78-CV-316-8W/W, cover all but the lowest water i conditions expected at Byron and Braidwood. This lowest condition is somewhat lower in both temperature and pressure than that cf the tests, and is considered less severe, and bounded by the water tests. .While the possibility exists for COMS PORV actuation under steam conditions, these are similar to, but less severe than normal PORV actuation. This case l is bounded by the Marshall steam tests, which were steam at normal plant operating pressure and temperature. Therefore, valve operability for COMS actuation is bounded by existing EPRI test data that demonstrated satisfactory operation. O
Pege'6 - Attachm nt A. Byron /Braidwood. Plant Specific Responses to NRC Ouestions 'gIgerding NUREG-0737. II.D.1 A ) 5. Bending Moments On Valves (Byron 1 and 2; Braidwood 1 and 2) The Licensee did not address the effect of bending moments induced on the valve inlet and discharge flanges on operability of the safety valves and PORVs. To evaluate the operability of the valves under the action of the maximum bending momente, the worst case loading must be considered. This loading should include the effects of dead weight, thermal expansion, earthquake (SSE), and valve actuation loads. Provide a comparison of the maximum predicted bending moments on the safety valves and PORVs and the moments applied to the test valves to demonstrate that the valves will perform satisfactorily under the predicted.ending moments. Response (Byron 1 and 2; Braidwood 1 and 2) Maximum bending moments for the safety and relief valves were calculated for . Byron /Braidwood including dead weight, the worst thermal expansion case, SSE 'loadir.g, and the maximum valve thrust loads. The moments were calculated for each safety valve flange and relief valve nozzle. The largest bending moment calculated for each valve type is presented below. For comparison, the largest moments applied during the EPRI tests are alss provided. The safety valve moment is that induced in test 908. The relief valve moment is from test 64-CV-174-23. The Copes-Vulcan valve with 17-4 PH '(-~) internals was the only configuration tested with moments, as it was assumed that the change in plug material would not 'r?t the results of the end load tests. Byron /Braidwood EPRI Maximum Maximum Bending Moment Pending Moment Eafety Valve 177,460 in-lbs. 298,750 in-lbs. Relief Valves 65,860 in-lbs. 43,000 in-lbs. The plant moment is less than that of the EPRI tests for the safety valve; thus the test demonstrates the valve operabiltiy. The moment calculated for the Byron and Braidwood relief valves, however, exceeds the moment tested by EPRI. Although the test cannot be used to demonstrate operability, these loads have been qualified by analysis. This maximum loading results in stresses cf 71% of the allowable value. The analysis is available for inspection at the Westinghouse Electric Corporation offices in Churchill, PA.
l P::ge 7 - Attachm*nt A. Byron /Braidwood. Plant Specific Responses to NRC Oucstions R?nerding NUREG-0737. II.D.1 qb 7. PORV Control Circuitry (Byron 1 and 2; Braidwood 1 and 2) l NUREG-0737, item II.D.1 requires that the plant-specific PORV control circuitry be qualified for design-basis transients and accidents. The Licensee should provide information which demonstrates that the above requirement has been fulfilled. The Licensee may either provide documentation to confirm that the equipment has been qualified under 10CFR50.49, or submit the following material for a complete review of the qualification of the control circuitry under NUREG-0737: l a. Provide a list of all PORV control circuitry needed to mitigate NUREG-0737 transients such as the following: (1) Switchgear (2) Motor control centers (3) Valve operators and solenoid valves (4) Motors (5' Logic Equipment ab) Cable (7) Connectors (8) Sensors (pressure, pressure differential, temperature, flow and level, neutron, and other radiation) (9) Limit Switches (10) Heaters (11) Fans (12) Control Boards (13) Instrument. racks and panels (14) Electric penetrations (15) Splices (16) Terminal Blocks b. For each item of equipment identified in 1, provide the following: (1) Type (functional designation) (2) Manufacturer (3) Manufacturer's type number and model number '4) Plant ID/ tag number and location c. For each item of equipment listed above, provide the environmental envelope, as a function of time, that includes all extreme parameters, both maximum and minimum values, expected to occur during NUREG-0737 transients, including postaccident conditions, d. For each item of equipment identified above, state the actual qualification envelope simulated during testing (defining the duration of the environment and the margin in excess of the design requirements). 1 If any taethod other than type testing was used for qualification, identify the method and define the equivalent " qualification envelope" so derived. 1 l l I
f l l l Page 8 - Attachm?nt A. Byron /Braidwood. Plent Specific Responses to NRC Questions Regerding NURCG-0737. TI.D.1 7. PORV Control Circuitry (Byron 1 and 2; Braidwood 1 and 2) Cont'd. e. Provide a summary of test results that demonstrates the adequacy of the qualification program. If any analysis is used for qualification, justification of all analysis assumptions must be provided. f. Identify the qualification documents that contain detailed supporting information, including test data, for items D and E. Response (Byron 1 and 2; Braidwood 1 and 2) The PORV and the necessary control circuitry and support systems to ensure its operability during postulated transients and accidents are comprised of Safety Category I equipment. As a result, these components are seismically qualified and, for electrical and active mechanical components located in a harsh environment, environmentally qualified in accordance with 10CFd50.49. The components appear on the Byron and Braidwood Safety Related Component Lists (SRCL) and have been addressed in the Byron /Braidwood Equipment Qualification Program. This program has been reviewed by the NRC and found to be acceptable as documented in the Byron Safety Evaluation Report (NUREG-0876, Supplement No. 5) and the Braidwood Safety Evaluation Report (NUREG-1002, Supplement No. 2). l 0
Page 8 - Attechm*nt A. Byron /Braidwood. Plant Specific Responses to NRC Ouestions Rencrding NUREG-0737. II.D.1 r-Question 8 (Byron 1 and 2; Braidwood 1 and 2) The Licensee stated that based on a probabalistic risk study of high pressure liquid challenges to safety and relief valves at Byron /Braidwood, the frequency of occurrence of water discharge was extremely low. Therefore, it was concluded that water discharge through safety and relief valves need not be considered in the valve evaluation. However, the Westinghouse Valve Inlet Fluid Conditions Report -hows that in an analysis of the Byron /Braidwood plants, the safety and relief salves opened on saturated steam at about seven minutes into the transient followed by a transition to saturated liquid discharge at thirteen minutes into the event. This indicates that water discharge through the valve is possible. Also, NUREG-0737 specifically requires that the safety valves and PORVs be qualified for inlet fluid conditions resulting from transient and accidents referenced in Regulatory Guide 1.70, Rev. 2. The feedwater line break accident must be analyzed, even though the probability is low. The EPRI test results demonstrate operability of the Byron /Braidwood safety valves and PROVs for a feedwater line break at Byron /Braidwood. However, the fluid loads due to the liquid discharge through the safety valves and/or PORVs must be considered in the analyses of the safety valve /PORV piping system.
Response
This study was an exhaustive and in depth assessment of the high pressure liquid challenges to the safety and relief valves. The study performed a plant specific evaluation including operating procedures. The Westinghouse Valve ()InletFluidConditionsReportwasnotplantspecific,didnotincludeoperator action, and included multiple system failures and therefore very conservative. The feedwater line break accident was analyzed, as well as other high pressure liquid challenges, and the analysis concludes that the probability of either saturated or sub-cooled liquid (excluding the loop seal) entering the safety valves cr PORV's was low. The analyses examined events included in Reg. Guide 1.70 Rev. 2 which showed the events and the subsequent challenges to the safety-valves to be on the order of 5 E-08 events per reactor year. u
Pcgs 9 - Attachment A. Byron /Braidwood. Plant Specific Responses to NRC Ouestions R*Rarding NUREG-0737. II.D.1 1 1 r Question 9 (Byron 1 and 2; Braidwood 1 and 2) The submittal did not provide detailed information on the thermal hydraulic and structural analyses of the safety valve and PORV piping. To allow for a complete evaluation of the methods used and the results obtained from the thermal hydraulic analyses, provide a discussion of the latest thermal hydraulic analysis containing at least the following information: a. Identification of any fluid transient cases analyzed that involve PORV discharge. The submittals do not mention that any transients other than safety valve actuations were considered in the analyses. The piping system must be analyzed for the most severe transient involving actuation of the PORVs. This could be liquid discharge following a main feedwater line break. l b. Identification of important parameters used in the thermal hydraulic analysis i and rationale for their selection. These include peak pressure and pressurization rate, valve opening time, and fluid conditions at valve opening. c. Verification that the loop seal temperature distribution assumed in the l j analysis is valid. d. Clarification on the codes used to generate the fluid forcing functions. The submittal stated that the Westinghouse ITCH code was used to calculate the forces. Normally, Westinghouse uses the ITCHVALVE/FORFUN combination of ('\\ programs to generate fluid pressures, momenta, and forces. Explain whether \\ this was the case for Byron /Braidwood. e. An explanation of the method used to treat valve resistances in the analysis. Report the valve flow rates that correspond to the resistances used. Because the ASME Code requires derating of the safety valves to 90% of actual flow capacity, the safety valve analysis should be based on flows equal to 111% of the valve flow rating, unless another flow rate can be justified. Provide information explaining how dorating of the safety valves was handled and describe methods used to establish flow rates for the safety valves and PORVs in the analysis, f. 'The sensititivity studies discussed in Appendix C of Reference 5 indicate that the calculated piping forces using RELAPS/ Mod 1 are quite sensitive to the node spacing and time step used in the thermal hydraulic analysis. For example, increasing the control volume size in pipe segments from 0.5 ft. to 1.0 ft. decreased the maximum forces in the segments by up to 50%. Similar node size and time step sensitivity could be a potential problem with other thermal hydraulic codes. Thus, provide the following: (1) Time Step: Because this is a parameter which can affect the calculation of piping forces, provide the range of time steps used. Also the time step used must be small enough to accurately calculate the propagation of the chock wave as it travels down the discharge pipe. Based on EG&G Idaho experience, the criterion for maximum time step size should be (~ 6t 21Ax/en, where c = sound speed, o,x = minimum node length, and n is a \\ multiplier that varies depending on the pipe fluid conditions. Since
l ~ l Page 10 - Attachm?nt A. Byron /Braidwood. Plant Specific Responses to NRC l-Ousstions Rezerding NUREG-0737. II.D.1 0 f.(1) Cont'd, ) l the shock wave could travel'at twice the sonic velocity n should be at least 2 and might need to be.as large as 5 in subcooled liquid where c is approximately 5000 ft/s. Demonstrate that the time steps chosen l I satisfy this criterion. If larger time steps were used, justify that I the larger time steps do not invalidate the analysis, or redo the analysis with smaller time. Since the maximum time step size is l related to volume length, the response to part (2) below needs to be considered when answering this question. (2) Node Size: The node size selected must be small enough to prevent the piping forces from being underestimated due to numerical smearing. Appendix C of Reference 10 indicated node sizes as small as 0.5 ft. may be needed to accurately calculate the piping forces. Discuss the node size used in the Byron /Braidwood model. ' Provide evidence to show bounding piping forces were calculated with the nodalization. g. A sketch of the thermal hydraulic model showing the size and number of fluid control volumes, h. A ccpy of the contractors thermal analysis report. Response (Byron 1 and 2; Braidwood 1 and 2) O Attachments 1 thru 4 present detailed information on the thermal hydraulic analyses for Byron Units 1 cnd 2 and Braidwood Units 1 and 2. Responses for Question 9 are presented in Section 4.6 and Figures 4-1 and 4-2 of Attachments 1 thru 4. O
Pega 11 - Attechment A. Byron /Braidwood. Plent Specific Responses to NRC Ouestions Renerding NUREG-0737. II.D.1 Ouestion 10 (Byron 1 and 2; Braidwood 1 and 2) To allow for a complete evaluation of the methods used and results obtained from the latest structural analysis, provide reports containing at least the following information: a. Identification of the computer programs used for the structural analysis. Provide verification that the computer program has been benchmarked for similar water hammer discharge problems (e.g., provide comparisons of calculated results with EPRI test results). b. A sketch of the structural model showing lumped mass locations, pipe sizes, and application points of fluid forces, c. A description of methods used to model supports, the pressurizer and relief tank connections, and the safety valve bonnet assemblies and PORV actuator, d. Identification of the key input parameters used in the structural analysis such as the damping factor, the computation time step, and the highest structural frequency (cut-off frequency) considered in the analysis. Give the maximum mass point spacint used for each pipe size and justify that this modeling technique would ensure that all of the desired structural frequencies would be accounted for in the structural response. ( ) e. Presentation of load combinations used in the analysis for both the piping, and pipe supports. The stress analysis should be performed for normal operation, thermal expansion, seismic (OBE and SSE), and safety valve and PORV actuation loads. As a F1nimum, list all load combination equations used for the piping and pipe support evaluation and the applicable stress limit for each equation. These should include normal, upset, emergency, and faulted conditions. Provide the equations used for the piping and supports both upstream and downstream of the valves. Differentiate between stress limits used for piping and pipe supports. Explain how the loads were numerically combined (algebraic, absolute sum, SRSS, etc.). f. An evaluation of the safety valve and PORV piping stresses including at least the following items: (1) Identify the governing code used for the piping design such as ASME or ANSI codes. Give the specific edition and date of the code used (year issued, addenda, and revision dates, etc.). (2) Provide a comparison of the worst case stresses in the piping with the l allowable stresses for each service condition (normal, upset, emergency, faulted). The worst stresses should be given for selected representative piping components (straight pipes, c1 bows, tees, reducers, etc.) of various pipe sizes and materials. The data provided in the stress comparisons should include the structural node number, pipe component description, location of pipe segment (upstream or 'N downstream), the maximum calculated stress, the applicable allowable stress. l
page 12 - Attachment A. Byron /Brnidwood, plant Specific Responses to NRC Outstions R*R*rding NUREG-0737. II.D.1 tv gunstion 10 (Byron 1 and 2; Braidwood 1 and 2) Cont'd (3) Identify any overstressed locations and describe any modifications needed to achieve acceptable stresses, such as installation cf insulation. g. The Licensee should provide a summary of piping support stresses which includes but is not limited to the following: (1) Identify the stractural code used to establish allowable support loads. Explain how allowable support loads were established for normal, upset, emergency, and faulted conditions. 1 (2) provide a comparison of the support stresses with allowable stresses for each service condition (normal, upset, emergency, and faulted). Identify the support type and support locaticn. (3) Identify any overloaded supports and describe any modifications or additions to the supports. (4) provide sketches of the piping system showing support locations. Assure that copies of the sketches are clear and readable. ( } h. provide a copy of the structural analysis report. Response (Byron 1 and 2; Braidwood 1 and 2) Section 4 and Figures 6-1 through 6-4 of Attachments 1 thru 4 describe the structural computer programs, the benchmarking ef fort, the structural models, the modeling methods and the key input parameters for Byron Unita 1 and 2 and Braidwood Units 1 and 2 respectively. Section 2 of these attachments discuss the applicable code and the load 1 combination criteria. Stress summary tables are presented in Section 6 for the j highest stressed locations. There are no overstressed locations. Therefore. response to Question 10, with the exception of 10E and 10G, are provided in Attachments 1 through 4. l The responses to Questions 10E and 10G for Byron Station, Units 1 & 2 are provided below. ) E. The pSARV system consists of Class I piping / supports in the upstream segments and B31.1 piping / supports downstream of the valves. All Class 1 pipe supports were designed and evaluated to meet the requirements of Westinghouse Design Specification 955349, Revision 1. The loads considered in the design were deadweight, pressure, thermal, seismic, ("} and the valve thrust. The loading combinations are described in the tabic Nv/ in Attachment 7. i _ = _ _ _ - - - _ - _ _ _. _ _ _ _ _ _ -. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .a
page 13 - Attechment A. Byron /Braidwood. plant Specific Responses to NRC-Questions Regerding NUREG-0737. II.D.1 Question 10 (Byron 1 and'2; Braidwood 1 and 2) Cont'd .v For the Decign, Normal, Upset, Emergency, and Faulted loading conditions, the loads indicated in the table were added algebraically, except as noted otherwise in the table in Attachment 7. The structural code governing the upstream support design is the ASME Boiler.and pressure Vessel Code Section III Subsection NF (1977 Edition through 1979 Summer Addenda). The upstream supports (identified as Class A in Attachment 8 were designed by considering service level loading conditions and the applicable service level stress allowable. The allowable stress limits for linear type supports are defined by the Code based on a working limit for design and service Level A (called normal operating condition in Attachment 7). Appropriate increases are identified for Levels B, C, and D; these are called out as upset, emergency and faulted, respectively. Stress limits for the evaluation of Level D conditions are specified by Appendix F of the Code. For the NNS downstream supports (identified as Class D in Attachment 8) also, the support loads for the normal, upset, emergency and faulted loading conditions were combined according to the table in Attachment 7. The support stresses for the normal and upset loading conditions were evaluated against the Service Level A stress limits. The stresses for other loading conditions were limited to faulted load allowable stresses. G(1) The design and analysis of piping supports was performed in accordance with the requirements of ASME Boiler and pressure Vessel Code (1977 Edition through Summer 1979 Addenda), Section III, Subsection NF. The allowable stress limits for linear type supports are defined by the Code () based on a working limit for design and Service Level A (called normal operating condition in Attachment 7). Appropriate increases are identified for Levels B, C, and D; these are called out as upset, emergency and faulted, respectively. Stress limits for the evaluation of Level D conditions are specified by Appendix F of the Code. G(2) A summary of stresses in support structures is provided in the tables in. In these tables, the ratio of the actual stress to the allowable stress is provided for the most severely stressed structural component (member or weld) for each support structure. The tables also indicate whether the support is a snubber, rigid, or a constant spring type' support. The locations of the supports are shown in Attachment 9. G(3) There are no overloaded supports, and no modifications or additions are planned at this time. G(4) The support locations are shown in Attachment 9. The responses to Questions 10E and 10G for Braidwood Station, Units 1 & 2 are provided below: E.&G. The piping design of Braidwood 1 and 2 is a replicate of Byron 1 and 2. Although the piping supports for Braidwood 1 and 2 differ from those for Byron 1 and 2, the Braidwood supports have been qualified in accordance to h the same structural code requirements as the Byron supports, utilizing the V same load input data (including loads due to safety valve and FORV actuation and slug discharge). The support calculations meet all applicable code and licensing commitments.
l { l pegs 14 ~ Attachm-nt A. Byron /Braidwood, plant Specific Responses to NRC Ounstions Regerding NUREG-0737. II.D.1 t \\ ()\\ Question 11 (Byron 1 and 2; Braidwood 1 and 2) According to results of EpRI tests, high frequency pressure oscillations of 170-260 Hz typically occur in the piping upstream of the safety valve while loop seal water passes through the valve. An evaluation of this phenomenon is l documented in the Westinghouse report WCAp 10105 and states that the acoustic l pressures occurring prior to and during safety valve discharge are below the maximum permissible pressure. The study discussed in the Westinghouse report datermined the maximum permissible pressura for the inlet piping and established the maximum allowable bending moments for Livel C Service. Condition in the inlet piping based on the maximum transient pressure measured or calculated. j While the internal pressures are lower than the maximum permissible pressure, the pressure oscillations could potentially exite high frequency vibration modes in the piping, creating bending moments in the inlet piping that should be combined with moments from other appropriate mechanical loads. provide one of the following: (1) a comparison of the expected peak pressures and bending woments with the allowable values reported in the WCAp report, or (2) justification for other alternate allowable pressure and bending moments with a similar comparison with peak pressures and moments induced in the plant piping. As part of the response to this concern, provide a description of the safety valve pipe inlet configuration (pipe size, length, and material). () Response (Byron 1 and 2; Braidwood 1 and 2) The evaluation of the pressure oscillation phenomenon while the loop seal water I passes through the safety valve is addressed in Section 6.1.1 of Attachments 1 thru 4 for Byron Units 1 and 2 and Braidwood Units 1 and 2 respectively. l l i l l 1 l l .}}