ML20215B269

Forwards Response to NRC 870108 Request for Addl Info Re NUREG-0737,Item II.D.1, Performance Testing of Relief & Safety Valves. Repts on Verification of Pistar,Gensap, Baseplate II & Stardyne Computer Programs Also Encl

ML20215B269
Person / Time
Site:	Arkansas Nuclear
Issue date:	06/12/1987
From:	Enos J ARKANSAS POWER & LIGHT CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20215B273	List: ML20215B269 ML20215B439 ML20215B465 ML20215B508 ML20215B530 ML20215B545 ML20215B568
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.1, TASK-2.B.2, TASK-2.D.1, TASK-2.D.3, TASK-2.F.1, TASK-2.F.2, TASK-TM 1CANO68702, IEB-79-01B, IEB-79-1B, IEB-85-003, IEB-85-3, NUDOCS 8706170276
Download: ML20215B269 (17)

Text

{{#Wiki_filter:y ..n, ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551 LITTLE ROCK. ARXANSAS 72203 (501) 371-400C June 12, 1987 l 1 l 1CAN068702 U. S. Nuclear Regulatory Commission { Document Control Desk Washington, D.C. 20555

SUBJECT:

Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 NUREG-0737, Item II.D.1, Performance Testing of Relief and Safety i Valves, Request for Additional Information Gentlemen: Your request, dated January 8, 1987 (ICNA018702), for additional information j concerning NUREG-0737, Item II.D.1, Performance Testing of Relief and Safety i Valves, is responded to in the attachment to this letter. I Very truly yours, y$ f ,/ ' J. Tedlno, Manager g e f Nucl6ar E gineering and Licensing e JTE:lw Attachment Enclosures 1 cc: Mr. Robert D. Martin Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 l i Mr. Robert Johnson y Resident Inspector Arkansas Nuclear One 0jI Russellville, AR 72801 t 8706170276 870612 ) PDR ADOCK 05000313 P PDR MEMBEA MCOLE SOUTH UTiUTIES SYSTEM

ARKANSAS POWER & LIGHT COMPANY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION DATED JANUARY 7, 1987 (1CNA018702) NUREG-0737, ITEM II.D.1 k PERFORMANCE TESTING OF RELIEF AND SAFETY VALVES FOR ARKANSAS Ntl CLEAR ONE UNIT 1 i i l 1 i i ATTACHMENT LETTER NUMBER 1CAN068702 JUNE 5, 1987

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION QUESTION 1 Dresser Ind., in March 1976, recommended to Metropolitan Edison Co. that the PORV block valve be closed at pressures below 1000 psig to prevent steam wirecutting of the PORV seat and disk. Testing by Dresser later showed.the 1000 psig pressure limit to be overly conservative and that the PORV asi designed was. qualified to system pressures of 100 psig. Below 100 psig the deadweight of. the lever on the pilot valve was sufficient to keep the ' pilot valve open. Dresser recommends, if the plant is to operate at pressures below 100 psig, that heavier springs be used under the main and pilot disks to ensure closure. Without the heavier springs recommended by Dresser, the PORV should not be used at system pressures below 100 psig. Since the minimum operating pressure at ANO-1 is below 100 psig, verify that AP&L has installed the heavier springs in its PORV consistent with Dresser recommendations.

RESPONSE

Discussions with Dresser engineering personnel indicated that the heavier springs are now supplied as standard replacement parts. Efforts are currently underway to procure the new springs for both pilot and main disks for installation during the next Unit I refueling outage. QUESTION 2 Provide the torque setting for the PORV block valve operator at ANO-1 and the torque produced at this setting. If the torque is less than 82 ft-lbs (the minimum torque tested by EPRI), it is the staff position that it is not adequate to conclude proper operation solely on manufacturer's calculations. The problems encountered with Westinghouse gate valves on closing, which were traced to the calculations used to size the valve operator torque requirements, indicate the need to experimentally verify the adequacy of the block valve / operator combination. AP&L should provide test data to demonstrate the SMB-00-10 operator at ANO-1 is-capable of providing adequate torque to close the block valve. 1 h

l

RESPONSE

i I i The ANO-1 PORV block valve is a Velan Engineering 2 " gate valve. Required opening and closing thrusts / torques were calculated for this valve and tested with M0 VATS equipment consistent with I.E. Bulletin 85-03 requirements. Testing indicated torques and thrusts were more than adequate to perform the intended function. 1 During the most recent refueling outage the torque switches on the limitorque SMB-00-10 operator for the block valve (CV-1000) were set as follows: Torque Switch Torque Switch Setpoint Trip (ft-lbs) Open Torque Switch 2.75 181 Close Torque Switch 1.50 141 This test data demonstrates that the PORV block valve operator is capable of providing adequate torque to close the block valve. QUESTION 3 Provide results of RELAP5-FORCE verification calculations for EPRI/CE tests for our review now and the complete verification document, discussed in your response to our request for information, when finished by UCCEL.

RESPONSE

The recently released RELAP5-FORCE verification document prepared by Gilbert Associates, Inc. for UCCEL is being provided as an enclosure. QUESTION 4 Insufficient detail was received on the key parameters used in the RELAPS-FORCE thermal-hydraulic analysis. Provide information on the node size, time step size, and choked flow location used in the RELAP5-FORCE analysis.

RESPONSE

The node size used was l'. The time step was 0.0002 seconds. l The valve opening time was 0.015 seconds for the PSVs and 0.06 seconds for the PORV. j Choked flow was assumed to occur in the relief val'.es. 2

[ i L l i -It is standard procedure with a program such as RELAP to perform a nodalization study to determine the number of control volumes required to obtain an accurate solution. This study usually takes the form of rerunning ] the model with more and more control volumes until the answers no longer-l change significantly. This type of analysis was performed by EPRI and the conclusion was that a minimum of five (5) control volumes per straight segment of pipe was required to obtain proper force resolution. Using the EPRI criteria, the number of control volumes required to model the ANO-1. ) safety / relief valve discharge piping system would greatly exceed the capacity of the available RELAPS program. This required: 1. Modeling the ANO-1 Safety Relief Valve discharge as three separate systems. One for each valve (two Safety Valves and one PORV). 2. A nodalization study to develop force intensification factors, accounting for the coarse nodalization. 1 Even when the ANO-1 discharge system was divided into three (3) separate .q systems, there still was not sufficient control volume available'in the q RELAP program. This required the application of force intensification j factors on those pipe segments with less than five (5) control volumes. The intensification factors were applied as. multipliers to the tabulated force time history when these force time histories were input to the structural program. QUESTION 5 The information supplied by AP&L in response to our request for additional information indicated a number of NUTECH and CDC programs were used to analyze the piping and supports. Provide information on the verification of PISTAR, GENSAP, BASEPLATE II, and STARDYNE and comparisons of calculated results to EPRI/CE data. RESPON>E i This response addresses computer programs used for the structural analysis j of ANO-1 piping and pipe supports. To simplify review, portions of our May 7, 1985 report (OCAN058505) are repeated here. The ANO-1 Pressurizer Safety and Relief Valve discharge piping was analyzed for the loads resulting from actuation of the PORV and the Safety Valves using the PISTAR computer program. PISTAR is a proprietary piping analysis computer program owned and maintained by NUTECH, Inc. (NUTECH). PISTAR performs the analysis and evaluation of ASME Section III and ANSI B31.1 piping systems for static and dynamic loads. The analytical solvers used in PISTAR are based on the well known public domain program SAPIV, developed by the University of California at Berkeley. 3 i

i .I PISTAR has been verified in accordance with a quality assurance program which conforms to the requirements of 10CFR50, Appendix B, ANSI N45.2.11 as amended by the USNRC Regulatory Guide 1.64, and Section III of the ASME Code. The verification report for PISTAR is provided as an enclosure. The verification was performed by comparing the important portions of the PISTAR solution for a series of benchmark problems to that obtained from manual calculations or from other computer programs such as ANSYS and EPIPE. Results of these comparisons showed good agreement between PISTAR and the manual calculations and other computer programs. The analysis of the piping for the SRV discharge loads was performed using ) the direct integration time-history solution technique. This solution technique is commonly used in determining response of structural systems to impulsive type loads such as SRV discharge loads. The validity of using this technique for this application has been previously demonstrated by comparison to actual test results. The solution technique in PISTAR was verified as described above. J The piping supports were analyzed for the support reactions obtained from j the piping analysis. The supports were analyzed statically using manual techniques and using computer programs. The computer programs used in ) performing the analysis of the supports are public domain programs which are ] used extensively in the nuclear industry by a variety of utilities, A/E's j and consultants for safety-and non-safety related applications. These { 1 computer programs are described below: 1) GENSAP Computer Program -- GENSAP performs the static analysis of elastic structures. This computer program is available through Control Data Corporation (CDC) and has been verified in accordance with NUTECH'S Quality Assurance Program. The NUTECH verification documents are enclosed. 2) BASEPL. ATE II and STARDYNE Computer Programs -- BASEPLATE II is a preprocessor to the STARDYNE Computer Program and is used to generate the required input data for the STARDYNE subprograms STAR and SPRINGS. This combination of programs performs the non-linear flexible analysis of baseplates. BASEPLATE II and STARDYNE are available through CDC and are on the CDC nuclear safety related list of computer programs which are currently subject to COMS0VRCE/CYBERNET Quality Assurance Policies. This list is attached, and provides documentation and verification of these codes. The question being addressed by this response also requests comparison of calculated results of these structural analysis computer prograns to EPRI/CE data. We believe this question arises from a stated secondary ibjective of the EPRI/CE program, which was to "obtain-sufficient nioing thermal l hydraulic load data to permit confirmation of models wnich may be utilized in plant unique analysis of safety and relief valve discharge piping i systems." It was the intent of the EPRI/CE program that the utilities be able to use the thermal hydraulic code (RELAPS/ MOD 1) confirmed by EPRI on EPRI/CE test data or to be able to verify other thermal hydraulic analysis i' computer programs with EPRI test data. The EPRI report NP-2479, Final 4

?: Report, dated December, 1982, provides the EPRI verification for RELAPS/MODl. It is entitled " Application of RELAPS/ MOD 1 for calculation of Safety.ard Relief Valve Discharge Piping for Hydrodynamic Loads." AP&L provided thermal hydraulic transient analysis for ANO-1 using the RELAP5/M001 computer program as modified by Gilbert Associates, which is called RELAPS-FORCE. Verification of this thermal hydraulic computer program is provided as an enclosure, and includes comparison of the calculated results of the RELAPS-FORCE code to EPRI/CE test data. We have provided comparison of calculated results of the thermal hydraulic computer program to EPRI/CE data. We do not have plans to perform the qualification tests needed in order to compare the structural computer programs to EPRI/CE data and cannot determine an appropriate function for any such comparison of structural code results. As described in the EPRI Report " Application of RELAP5/ MOD 1 for calculation of Safety and Relief Valve Discharge Piping for Hydrodynamic Loads" EPRI NP-2479, Final Report, December 1982, the EPRI reports did not intend that utilities carry out plant-specific verification of structural analysis codes against EPRI data. Structural analysis was qualified during the EPRI/CE test program for the limited purpose of supporting verification of the RELAPS/M001 thermal hydraulic analysis computer code. If the EPRI/CE test facility and its support system could have been infinitely rigid, the measured test data would have consisted only of the hydrodynamic loadings. If this were the case, the hydrodynamic loads calculated using RELAP5/ MOD 1 analyses could have been compared directly with the recorded test data. The design approach established for the EPRI/CE test loop piping supports 1 was to provide supports which would facilitate experimental measurement of piping loads. Based on this design goal, extremely rigid dynamic support structures were designed for the test valve discharge piping, The test facility structure was not sufficiently rigid, however, to prevent the structural dynamic response characteristics of the facility to cause the recorded data to deviate somewhat from the hydrodynamic loading. J l The RELAPS/ MOD 1 predicted hydraulic loads, in ' he form of time history forcing functions, were used in the EPRI program as inputs to the structural I computer programs. The results of the structural analyses were then compared to the test data. Because this comparison was a primary method of evaluation of the RELAP5/ MOD 1 thermal hydraulic code results, the dynamic response characteristics of the test facility had to be shown to have been accurately reproduced analytically. The method of qualifying the dynamic structural model was to compare the natural frequencies of the test facility with those predicted by the dynamic structural analysis. Two types of test were performed. The first was a snap-back test which consisted of lifting the first discharge elbow with a crane. The second was an actual valve actuation test. 5

t Plant specific qualification testing of this nature.is'not re' commended by ~ i ~ EPRI. The-test facility piping support configuration was not designed to'- simulate operating plant structural dynamics, so a comparisonito the'EPRI/CE-test facility.has no appropriate function. The reason for the EPRI/CE; structural qualification of the test facility was limited to validation of i the thermal hydraulic code RELAP5/M001. There.is no recommendation for validation'of structural codes by EPRI..In fact, the EPRI. report states, at' page 1-5: There are many dynamic structural analysis computer codes whn.h 1 are available to perform this type'of dynamic analysis, The mathematical equations-and-techniques utilized by these. programs.are well established. These programs have been )erified. M by comparison with well-documented theoretical'and. experimental: 'j problems. There is,.therefore, a high level of confidence that' the computer codes correctly solve the problems to which.they are. applied. In light of the recommendations of the' EPRI/CE. test reports, the documents provided should fully and satisfactorily resolve NRC concerns with. respect-j to the verification of the structural analysis computer codes used for . 1 plant-specific analyses of ANO-1. 1 QUESTION 6 The lumped mass spacing used in the structural model was not discussed 'in the AP&L response to our request for information. Provide this information. Also, it appears that the damping values in the structural analysis of piping are related to the PORV and safety valve discharge. loads. If these l values are to be consistent with the requirements of the Regulatory Guide 1.61, they should be related to the intensity of earthquake rather than discharge loads. Please clarify,the above. O

RESPONSE

j The lumped mass spacing used in the structural model.was that required to s obtain accurate dynamic responses of the piping system up'to a 50-Hz cut-off frequency. To achieve this accuracy, the maximum spacing between lumped masses was conservatively limited to one-half the length of an equivalent-simply-supported pipe whose natural frequency is 50 Hz. i 1 ) I 6 i ' (. a

The damping'v'alues used in'the structural analysis were determined for each dynamic loading condition. considered. For the earthquake loading conditions ~ the damping values used were those defined in the FSAR, i.e., "3, O.5%' damping for both the design earthquake and the max,imum earthquake loadings. 1As the PORV and Safety Valve discharge loads were not previously considered for these' systems, damping values for.these loading conditions are not specified in the.FSAR. For these loading conditions, the damping

values:used were.obtained from NRC Regulatory Guide-1.61 and were based on the. diameter of the piping and'the anticipated stress levels for the loading-

'l conditions. Although AP&L is not committed to Reg. Guide 1.61 for ANO-1,-it was utilized for guidance. Therefore, 1% damping was used for the PORV -loading conditions and 2% damping was used for the Safety Valve discharge '? 'lcading condition, consistent with Reg.' Guide'1.61. QUESTION 7 The submittal stated the discharge ' loads due to valve actuation (PORV or SRV) were applied independently to.each discharge piping system. The. piping responses were then combined absolutely to obtain the total ~ system response. j Provide a more detailed description of how this was done. How was.the response of the common header upstream of the relief tank handled with this j method? Discuss how the effect of simultaneous (or nearly simultaneous) actuation of the valves was considered. Also a concern has been raised that combining the individual line responses absolutely could result in missing ) the critical frequency response of the total system. Address this concern in your response. Finally the submittal did not indicate which of the 1 transients analyzed was the limiting transient for the piping and support J stresses. Provide this information.

RESPONSE

I In the dynamic analysis of the piping system, the thrust loads on'the i discharge line from each valve are applied independently. Loads on the common header are considered, and analytically are included with one line. Responses from these three independent analyses are conservatively combined absolutely. This ensures that the responses obtained would bound any sequence of valve actuation, including the simultaneous actuation of all valves. As this is a linear, elastic analysis of the entire piping system using a conservative method of superposition of independent loads, the concern relative to missing critical frequency response of the system is not j valid. The limiting. transients were, for saturated steam, the rod ejection accident at hot zero power and, for subcooled water at 400 F, the steam line break q i accident. Additionally, the saturated water case was run on PSV 1001 for comparison but the force magnitudes proved to be less than those for-saturated steam and subcooled water, and therefore this case was not further considered as relates to limiting transients. 7 L

QUESTION 8 The load combinations used for piping support' analysis require explanation. l The staff position is that one of the'following alternatives should be used: (a) The load combinations contained in.the.FSAR with appropriate code allowable stresses or (b) The load comb %ations contained in the Standard Review Plan (SRP),. NUREG-0800, Section 3.8.4 "Other Seismic Category I Structures" with the appropriate. code allowable stresses. In any case, the load combination equations contained in the response to the Request for Additional Information, dated April 1984 do not address all of .l the conditions which should be considered [for example, a combination which l involves DE (design earthquake) is not included]. Indicate the manner in which you will ccmply with the above. l

RESPONSE

The load combinations contained in the response to the Request for ] Additional Information, dated April 1984, were only those combinations 1 including the new dis, charge loads. The load combinations considered in the 'nsidered in this analysis unless ] original qualifying analysis were they were bounded by a combinat: ' 3c.c 'g the new discharge loads. The j load combinations considered are J l DW + TE + DE < 1.0 times . bles DW + TE + RV < 1.0 times AISC allowables DW + TE + [(RV + SV)2 + ME2] g*.< 1.5 times AISC allowables

• As the ME earthquake bounds the DE earthquake, the combination of discharge loads with the DE earthquake is bounded by this combination.

] DW = Deadweight of the piping components and contained fluid (either saturated steam or subcooled water.) j TE = Thermal expansion of the piping and movement of the pipe anchors due to the expansion of the pressurizer and quench tank. DE = Design Earthquake ME = Maximum Earthquake RV = Discharge loads resulting from actuation of the PORV for either the saturated steam condition or subcooled water condition. 8

SV = Discharge loads resulting from actuation of-the safety valve for either the saturated steam cordition or subcooled water condition. These combinations meet the NUREG-0800, Section 3.8.4 standard. QUESTION 9 AP&L's response to our request for information stated that it was not clear what was meant by the qualification of the PORV coritrol circuitry. It was also stated that it was AP&L's understanding that it did not mean environmental qualification. Environmental. qualification, however, is exactly what was meant by the qualification of the PORV control circuitry in NUREG-0737, Item II.D.1. It is the staff position that in order to demonstrate the ANO-1 control circuitry is qualified to the requirements of NUREG-0737, the design qualificatior.s must be compared to the environment the control circuits will be exposed u. Provide documentation to show the PORV control circuitry has been qualified under 10 CFR 50.49, or to allow a complete review of the qualification of the control circuitry-for the PORV under NUREG-0737, provide the following: A. Provide a list of all PURV control circuitry needed to mitigate NUREG-0737 transients. Your list should include, as a minimum, the following: 1. Switchgear 2. Motor control centers 3. Valve operators and solenoid valves 4. Motors 5. Logic equipment 6. Cable 7. Connectors 3. Sensors (pressure, pressure differential, temperature, flow and level, neutron, and other radiation) 9. Limit switches 10. Heaters 11. Fans 12. Control boards 9 ) i

13. Instrument racks and panels 14. Electric penetrations 15. Splices ) 16. Terminal blocks B. For each item of equipment identified in A, 1. Type (functional designation) 2. Manufacturer 3. Manufacturer's type number and model number 4. Plant ID/ tag and 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, j both maximum and minimum values, expected to occur during NUREG-0737 transients, including post-accident conditions. D. For each item of equipment identified above, state the actual l qualification envelope simulated during testing (defining the duration of the environment and the margin in excess of the design l requirements). If any method other than type testing was used for qualification, identify the method and define the equivalent ) " qualification envelope":so derived. 1 l E. Provide a summary of test results that demonstrates the adequacy of the j 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. q

RESPONSE

The issue of environmental qualification has been thoroughly addressed by the NRC under 10CFR50.49. The determination was made that environmental qualification of PORV control circuitry was more than adequate considering a PORV system configuration which includes an environmentally qualified PORV block valve and environmentally qualified indication of PORV position. Such a configuration ensures the capability to detect and isolate a stuck open PORV. The mechanical safety valves provide the capability to protect the RCS from overpressurization, therefore the' capability to open the PORV is not essential. 10

i I In addition to this, our review of the licensing concern raised by this question shows.that the position we took on this issue (i.e., that "qualificatien" as used in NUREG-0737, Item II.D.1 does not imply " environmental qualification") in our May 7,1985 submittal (0CAN058505) is justified on several grounds. The express language of NUREG-0737, Item II.D.1 is: Reactor coolant system relief and safety valve qualification shall include qualification of associated control circuitry, piping and supports, as well as the valves themselves. The justifications for the position that " qualification" as used here does not imply " environmental qualification" include: NUREG-0737 clearly indicates when " environmental qualification" is required either by direct reference.to Appendix B environmental qualification criteria (c.f.: II.F.1-3, II.F.1-4, II.F.1-5, II.F.1-6, j and II.F.2) or by direct reference to the environmental qualification q Memorandum and Order CLI-80-21 (c.f.: II.B.1, 11.8.2, and II.D.3). However, these references and terminology were not used relative to PORV control circuitry. The terminology " environmental qualification" refers to a requirement of testing and documentation relating to certain electrical equipment exposed to potentially harsh environments. The word " qualification" here was used to describe the Item II.D.1 requirements relating to strictly mechanical components (e.g.: safety valves, piping, and supports) as well. This raises the presumption that " environmental qualification" was not intended. The NRC staff has not interpreted " qualification" as meaning " environmental qualification" in NUREG-0737 over the past seven years: 3 ) In 1980, the NRC staff issued to AP&L a satisfactory preimplementation review of the TMI Action Plan on this issue. (March 10, 1980, BCNA038051). The NRC staff accepted participation in the EPRI functional testing as completely and j satisfactorily addressing the requirements of this Item. The acceptance was conditional only on subsequent approval of the EPRI test program, which was satisfactorily given af ter its completion 3 (SECY-83-270, July 5, 1983). Environmental qualification of i control circuitry was not raised as a requirement. In October 1980, the NRC issued the Bulletin 79-01B. It requested a list of plant equipment which was to be environmentally j qualified under the requirements of the TMI Action Plan. (0CNA198083). AP&L responded, and its list did not include PORV control circuitry as equipment required to'be environmentally qualified (ICAN018110, January 14, 1981). AP&L's submittal remained unquestioned in this respect after review by the NRC. 11

The NRC staff has taken the position that'" qualification" does not imply " environmental qualification" of PORV control circuitry. Tae staff issued this position as NRC policy in SECY-83-270 (July 5, 1983). This policy statement specifically addressed NUREG-0737, Item II.D.1, the EPRI testing program, as well as concerns which had arisen subsequent to the TMI Action Plan relating to instrumentation and control of pressurizer relief and safety valves. The policy statement said that post-TMI valve instrumentation and control concerns would be addressed under the then newly issued 10CFR50.49. The policy statement concluded that the EPRI tests confirmed basic valve capa"i"ty. This position has authority because it was the interpretatlo: C o contemporaries of the NUREG who were in a position to know me intended scope of the TMI Action Plan, because it was maintained for a period of years, and because it was issued as NRC policy. The requirements of I the NUREG were met by the confirmation, through the EPRI testing, of basic valve capability. The issue of environmental qualification at ANO-1 has been thoroughly addressed by the NRC under 10CFR50.49 in their review of several submittals of lists of environmentally qualified electrical equipment and in an NRC audit conducted at the plant. Notwithstanding the above, we have re-reviewed the function of the PORV control circuitry with respect to environmental qualification. The PORV at ANO-1 is a pilot operated valve in which the pilot is actuated by an electrical push-action solenoid. When the solenoid is energized, the PORV is actuated open. When the solenoid is deenergized, the PORV closes. The closed position of the PORV is its safety position. The following analysis demonstrates that the solenoid will not be energized as the result of electrical failure caused by a harsh environment. The solenoid is controlled by signals which originate from pressure transmitters located in the reactor building. These signals are processed through a signal converter in the control room and the converted signals are fed to a bistable switch, also in the control room. The signal across the contacts of the bistable switch actuates the solenoid. The three potential failure modes which could be caused by a harsh environment are addressed as fellows: 1 (1) The failure of a pressure transmitter or associated circuit (including cable connectors and splices) is precluded by the fact that the pressure transmitters in question are environmentally qualified for long term post-accident conditions. (These pressure transmitters are associated with the Reactor Protection. System). (2) The failure of the solenoid would result in its deenergization. Deenergization of the solenoid results in the PORV closing, which is its safety position. (3) The failure of the solenoid-associated power circuit could occur by a short to ground, cable failure, or cable breakage. In any of these events the solenoid would deenergize. Deenergization of the solenoid results in the PORV closing, which is its safety position. i 12

It is, therefore, concluded that the PORV will not inadvertently open or remain open as a result of an electrical failure caused by a harsh environment. The harsh environments which are addressed by 10CFR50.49 will not affect the ANO-1 plant-specific control circuitry for the PORV in its ability to ensure that the PORV achieves and remains in its safety position. Therefore, we have demonstrated, in specific response to this question, that we are able to document an appropriate resolution of the issue of PORV control circuitry and its environmental qualification under 10CFR50.49, i i J l 13 ]

v. e ENCLOSURE 1 RELAPS-FORCE VERIFICATION REPORT i

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. o VERIFICATION OF THE RELAPS-FORCE l [ I E HYDRAULIC FORCE i l-CALCULATION CODE \\ I BY l J. M. CAJIGAS d GILBERT ASSOCIATES, INC. 4 8 READING, PA 19603 ) May 1984 e L l_ M M 3680 i J

.,i;..- 1.0-INTRODUCTION' RELAPS-FORCE (1) is a modified version of RELAP5/ MOD 1(2) which includes.a hydrodynamic forcing function calculation option. This version generates r time-dependent force functions for piping segments defined by the user. g.. RELAP5/ MOD 1 has been modified to solve the hydrodynamic force' equation for the requested RELAP5 volumes, at each time step, and write the resultant force to the RELAPS output print and plot' files. I This report documents and verifies the accuracy and validity of the changes to RELAP5/ MOD 1. The verification process will include: i

1) RELAPS/ MOD 1 Changes verification. This verification will show that the

.RELAPS-FORCE. modifications have not adversely altered the precision of the.RELAP5/ MOD 1 calculation.

2) Hydraulic Force Calculation Verification - EPRI/C-E PWR SRV Tests. This a

verification will show the adequacy and accuracy of the RELAP5-FORCE' { force calculation methodology by comparison to test data from the EPRI/C-i E PWR SRV Test Program (3).

3) Hydrualic Force Calculation Verification - Edwards' ~& Hanson's Pipe Experiments. This verification will show the adequacy and accuracy of the RELAPS-FORCE force calculation methodology.by comparison to test data I

reported by A. R. Edards(4) adn G. H. Hanson(5). This data is I particularly significant because it allows a better. verification of the' blowdown force option of RELAPS-FORCE than the one permitted 'from the. ~ l EPRI/C-E PWR SRV Test configuration and data. 4 w ~,' _b

2.0 RELAP5/ MOD 1 CHANGES VERIFICATION The RELAPS-FORCE code was developed by programming the hydraulic force equation into the RELAPS/ MOD 1 code. This required the addition and modification of subroutines to the program. 4l To rnow that these modifications did not alter the basic RELAP5/ MOD 1 Li calculations, a sample problem was run for the following two cases. I*

1) RELAPS-FORCE USER'S GUIDE (1) sample problem run without the force option cards using RELAPS/ MOD 1, Cycle 14 AW. RELAPS/ MOD 1, Cycle 14 AW, is the L

base code for RELAPS-FORCE, Version 14.

2) RELAP5-FORCE USER'S GUIDE sample problem run with the force option cards i

using RELAPS-FORCE, Version 14. m Figure 2-1 shows a schematic of the RELAPS-FORCE USER'S GUIDE sample problem. The RELAP5/M001, Cycle 14 AW and RELAPS-FORCE, Version 14 input i i listings are shown in Appendix B. {t t For comparison purposes, a portion of the output listing for the period between 0.04 sec and 0.0449 sec of each sample run is presented in Figures e i 2-2 and 2-3. Note that the listed results agree exactly. Therefore, it can h be concluded that the RELAPS-FORCE modifications to RELAPS/ MOD 1 have not i altered the internal thermal-hydraulic code calculations. i L a

• s s

.x r 200 400 1 301 501- [ k 301 501 "g 100 g P, X n i 502 j 302 F, n E n 8 o 1 303 P\\PE SOA 1 n 503 30 505 n I 600 { \\ v F' I 504 700 / B00 0 8 w l c_. c. 4 1001 o F. 1200 / 1003 l 1004 P, X, 1002 1003 NOTES: 1100

1. The length of pipe component v<>lumes = 1 ft.

~

2. Tee branch component volume length = 1 ft.

3. P1 = 1000 PSIA: P2 - 14.7 PSIA

5. Steam Quality = XI = X2 - 1.0

6. All piping 6" nomir.nl

~ (, F@EE. 2-1 ; RELAP6-FORE.E SA%4%. PtoBLE A u

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.V 9.. 1 3.0 HYDRAULIC FORCE CALCULATION VERIFICATION - EPRT/C-E PWR SRV TESTS ~ -3.1 The'EPRI/C-E PWR SRV Tests: E Under the management of the Electric Power Research Institute (EPRI)', a full scale PWR pressuri::er safety relief valve test program was I_. q carried out at'the' Combustion Engineering (C-E) test facilities in j Windsor, CT(3). The C-E test' facility was designed for full-flow-l tests of selected safety valves under a wide range of inlet fluid-conditions and inlet: piping configurations. Figure 3-1 shows an isometric of.the test facility piping. I Ons of the objectives of this program was to obtain sufficient piping load data to~ permit confirmation of analytical mode's. Thus,'the. test. I facility was equipped with instrumentation to record transient b parameters such as valve discharge line. pressures Land fluid induced loads. The EPRI/C-E PWR SRV Tests chosen for this verification are:

1. Test No. 1411: Steam Discharge

2. Test No. 908:

Cold Water Loop Seal. Discharge

3. Test No. 917:

Hot Water Loop Seal Discharge ( These particular tests were selected because they all used the same valve model (Crosby 6M6) and piping configuration. In addition, good 6 quality test data for these tests is readily available. The test data , I ~ L 1.- .)

,3 r i n is sunmari::ed in Reference 4, " Measurements of Piping Forces in a Safety Valve Discharge Line' included herein as Appendix A. j q d i I F 1 j l' 'l L E .. j s - ) I i-l t i I L4 c

{ e. p i i I f TEST VALVE a Sch. X1 i 1 8"Sch.16

1. SchAO f

_ f(% VENTURI g e ^ rg ~ TANK l 'i 12" Sch. 80 { t ^ 1. f~ RUPTbRE DISC. ASSEMBLY g Y* i 4 4 %)6u2f 5-t*. EPET./C.-E pwR ScV

• E5'i~ T-ACLtTy PLP46. I5bME"Eic.

L .)

i;. 1 . p.. 1 3.2' RELAPS-FORCE.MODEL OF'THE EPRI/C-E CROSBY 6M6 TEST FACILITY-Appendix A, Figure 3 shows a detailed drawing of the test _ facility-including the location of the. process instrumentation. Force' ' measurements were made by summing the output of a pair of. load cell. strain gage ' transducers per pipe ' segment (WE 28 throu'gh WE 35). [ a The RELAPS-FORCE model of the test configuration is shown on Figure 3-L 2. Note that the-nodal points for the pressure transducers used for-i' .the verification have'been' labeled in add'ition to the location and' direction of. the hydrodynamic piping loads ' calculated. A11 the' Crosby. 6M6 tests were performed with a loop-seal inlet. piping configuration. The verification approach will be to compare _the analytical results with the tests pressure and load measurements. Pressure comparisons j are used to verify the accuracy of the RELAP5/ MOD 1 code -and model in reproducing the test configuration physical geome'ry and transient t } thermal-hydraulic phenomena and thus adds further support to the force calculation validation. L s a-t L

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• t 4

s = 8 3 5 ], 4 l-E w o a 7 T 6

s. '

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• h

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• w a

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• c6*f e s"~o T

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f 3.3 COMPARISON BETWEEN RELAPS-FORCE CALCULATIONS AND TEST DATA e 3.3.1 Test No. 1411, Steam Discharge Tests No. 1411 simulates a continuous steam discharge through a PWR pressurizer safety valve. The valve inlet pressure was regulated by modeling the reservoir pressure to ramp from 2410 psia to 2540 psia in 0.5 see as indicated by Appendix A, Figure 6. The Crosby 6M6 valve j j 1 used for these tests had a full-open area of 0.0253 ft. However, an 2 d area of 0.0204 ft2 was used in the RELAPS model to achieve the test i measured steady-state steam flow rate. As indicated in Reference 4, 1 r this valve leaked slightly prior to the test and thus the initial a { downstream air was replaced with steam. Assuming constant enthalpy j thrcctling, a quality of approximately 0.90 is calculated for the l i downstream piping steam environment. Therefore, the RELAPS model k i downstream conditions for this case correspond to 0.90 quality steam j at atmospheric pressure and the pipe wall temperature initialized at 2120F. The valve opening characteristic model used for this test is depicted against test data on Figure 3-3. The full-opening time used was 15 msec. See Appendix B for a listing of the RELAPS-FORCE input for this transient. Figures 3-4 through 3-6 compare the RELAP5 calculated pressures with test measured data for three discharge piping pressure transducers, PT 9, PT 10, and PT 11 as shown on Figure 3, Appendix A. Considering the possible differences discussed below, the calculated pressure histories are in good agreement with the test data. hum Y

if The hydrodynamic piping forces calculated by RELAP5-FORCE for test No. 1411 are compared with the test data on Figures 3-7.through 3-10.-- It can'- be observed that the magnitude and timing of the RELAPS-FORCE calculated forces agree reasonably well Sith the' test data'. .A notable discrepancy, occurs near the 200 msee point where test' data for Forces 3 and 4 .f' indicates force peaks not reproduced.by the' code. -This difference is apparently due to the accumuletion of condensate:in the lower horizontal discharge piping leg prior to the valve opening. Althot.gh, as ' discussed above, an attempt was made to model the downstream steam environment, information on the accumulation of ' condensation in the discharge piping was not available to allow reasonable modeling of this condition. During. the test, the accumulated condensate was apparently collected. into a slug by the steam discharge producing the 0.2 see force peaks-in pipe. segments 3 ana 4 as the slug of water accelerated out of these pipe sections. An important aspect to be considered when the RELAPS-FORCE calculated loads are compared with test data is.that the test facility pipe supports stiffness was not high enough at some supports to allow a rea'sonable one-on-one comparison. Appendix A, Figure 5 indicates that measured loads on~ pipe segments 3 and 4 should be expected to be somewhat below the actual L support applied load due to low supports stiffness. In this regard,.It. i should be indicated that the force comparisons herein are intended to 'l verify the adequacy of RELAPS-FORCE for hydrodynamic force. calculations. I They are not to be considered as a substitute for the piping structural analysis required to better reproduce test load data for pipe segments with low supports stiffness. s 'b m

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a-t..- - ) 3.3.2 Test No. 908: Cold Water Loop Seal Discharge I Test No. 908 was performed with.the loop seal piping-filled with cold water. The valve inlet' pressure was regulated by modeling the reservoir-to de-pressurice from 2690 psia to 2670 psia in 0.5 see as indicated by Appendix A., Figure 16.. For this test, a. 0.0834 ft2 ' orifice in::talled at the exit of' the 12" discharge pipe was modeled as a RELAPS. abrupt-area-change junction ef the same area. As in test No. i 1411,.a 0.0204 ft2 full-open area was used-in the RELAPS model to achieve the test measured steady state steam flow rate.. Appendix A, i Figure 15 shows the loop seal temperature distribution prior to valve. opening. During the initial. opening phase, this valve chattered appreciably for about 0.9 sec. before opening fully in about 15 msec. Therefore, the loop seal water was discharged into the outlet. piping prior to the valve's full opening. For this reason, tre RELAPS model includes the lcop seal water volume distribution.in P - outlet pipe and a 15 maec linear valve opening characteristic as shown on Figure 3-11. In accordance with Reference 4, the downstream piping initial conditions were set for 80cF r.ir at atmospheric pr' essure and an assumed relative humidity of 907.. See Appendix B for a listing of the RELAP5-FORCE input for this transient. Figures 3-12 through 3-14 compare the RELAPS calculated pressure histories with test data at the location of pressure tranducers PT.9, PT 10, and PT 11. The calculated pressures agree reasonably with the test. data with two exceptions. First, the PT 9 pressure test data ~ exhibits an off scale peak on or about 25 msee believed to be of d i mee e

11 0 gl t I spurious nature. However the calculated ' peak pressure at this location occurred much earlier.into the transient than the seemingly-valid test data indicates. This difference'is probably due'to.the. ~ initial discharge piping. loop seal distribution discussed above. q l Second, the RELAP5 pressure histories'at the locations of PT 10 and PT -11 agree well with test-data in the early and. latter parts of the 3 transient but exhibit aepressurization between 0.10~and 0.2'5' secs.- Close scrutiny of the' RELAP5 results indicatei that this is duelto al depressurization downstream of the vertical discharge pipe's area ~ 4 . change due to supersonic phasic velocities. This. behavior is produced-by the absence of choking at.this area' change during obvious choking ^ conditions, i.e., fluid velocities larger than propogation velocities. As discussed below, a choking model was not applied to this area change junction to prevent underestimating pipe segment 2 loads'during. l the discharge of this cold loop seal slug. As the loop seal water.is discharged out of the piping system, steady.' state steam choked flow at the discharge orifice plate recovers the PT 10 and PT 11 pressure to the test meast cad levels. The hydrodynamic piping forces ' calculated by RELAP5-FORCE-for ' test No. 908 are compared with test data on Figures-3-15 through 3-17 The magnitude and timing of the calculated loads for segments 1'and 2 compare reasonably with the test data considering the-loop seal location modeling assumptions required by the initial valve chatter. s The higher calculated segment 3' loads can be attributed to the' ~ relatively low stiffness of this segment's supports which decrease the test measured loads. The' segment 3 calculated loads, however, seem j ~l 1 j

1 o 1 I reasonable since the resulting forcing function is expected when compared to the calculated and measured loads on the stiffly supported set; ment 2. It should be noted that the choking model was not applied i to the expansion Junctions on pipe seEment 2 to prevent i underestimating the loads on this segment due to the lower acceleration of the loop seal slug as choking occurs at these f junctions. I 1 m L L W h M

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5 f 3.3.3 Tests No. 917: Hot Water Loop Seal Discharge I Test No. 917 was designed to simulate a hot water loop seal discharSe. I Similar to test no. 908, this valve chattered for about 0.65 secs before opening at 2650 psia in about 90 msec. Therefore, the RELAP5 model includes a 90 msec linear valve opening characteristic, Figure 3-18, and 0.5 see inlet pressure ramp from 2650 psia to 2720 psia corresponding to the reservoir history shown on Appendix A, Figure 26. A RELAP5 model valve full-open area of 0.0194 f t2 as required to match the steady state steam flow rate measured during the test. Per L Reference 4, downstream initial conditions were set for 80o F air at atmospheric pressure and an assumed relative humidity of 90'/,. See Appendix B for a listing of the RELAPS-FORCE input for this transient. Figures 3-19 through 3-21 compare the RELAP5 calculated pressure L histories with test data at the location of pressure transducers PT 9 and PT 10. The calculated pressure histories are in reasonable agreement with the test measurements. Although the peak pressure measured at PT 9 was higher than the RELAP5 peak, this maximum occured at the time the loop seal slug was passing this point which resulted 1 in oscillatory pressure measurements. However, the average of the PT 9 peak pressure oscillations is in good agreement with the RELAPS calculations. Agreement between PT 10 test data and RELAPS is good except for the period between 0.2 see and 0.5 see where a supersonic ~ velocity depressurization phenomenon similar to the one discussed in section 3.3.2 occurs. consistent pressure recovery starts at about 0.35 secs with steady state choked steam flow at the exit no::le. m e

? e I u The RELAPS-FORCE piping force histories for test No. 917 are compared with test data on Figures 3-22 through 3-25. The magnitude and timing of the calculated loads are in good agreement with the test data with some exceptions. The segment 1 force function behavior indicates an oscillatory period in the early part of the transient. These l oscillations, also shown in the PT 09 pressure data (Figure 3-21) are the result of the valve chattering observed during this test. This resulted in an oscillatory opening and closing pattern during the valve opening period. These oscillations were not modeled into the RELAPS valve opening model and thus the pressure and force 1 oscillations induced by this condition were not reproduced by the code. Slightly lower segment 3 measured loads are attributed to low ( l supports stiffness (see Section 3.3.1). A wave contribution positive peak at about 0.22 see was calculated for the segment 4 force history f but was not measured by the test instrumentation. A review of the RELAP5 results indicates that this behavior was expected since the wave force peak corresponds to the time of loop seal slug acceleration p out of the segment 4 section. The segment 3 force test data indicates that the slug was still intact as it left this segment and thus a segment 4 wave force behavior like the one calculated by RELAPS-FORCE is expected. 5 i I 1 l 1

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p. ( 4.0 HYDRAULIC FORCE CALCULATION VERIFICATION, EDWARDS'AND HANSONS' PIPE r-l L EXPERIMENTS l" 4.1 EDWARDS' PIPE FXPERIMENT 1 The pipe blowdown experimental data reported by Edwards (5) in 1970 provides an excellent experimental data base to benchmark the blowdown force option of RELAP5-FORCE. A schematic of the experimental facility is shown on Figure 4-1. Note the pressure of gauge stations (GS1 to GS7) used to measure the transient pressure, temperature, and u void fraction in addition to the load cell used to measure the hydrodynamic pipe axial load. e ( The experimental tests consisted of pressurizing the pipe with water i i i to the required test pressure and rupturing a glass disc at the end of l the pipe with a pellet gun to initiate the blowdown. It was observed that some of the glass dise was retained around the circumference of the disc support assembly reducing the discharge area by.as much as i 15%. The test used for this verification was initiated with water conditions of 1000 psig and 4670 F (saturation pressure : 285 psig) 4.1.1 RELAPS-FORCE MODEL OF THE EDWARDS' PIPE EXPERIMENT i 1 i The RELAPS-FORCE model of the test configuration is shown on Figure 4 4-2. RELAP5 output data requests (minor edits) were selected for } those volumes corresonding to the gauge station locations and for the pipe wave, blowdown, and total force calculated. i._ s J

1 i I The verification will consist of pressure, temperature, void fraction, f and load comparisons between RELAPS-FORCE and the test data. As d, indicated in Section 3.2, thermal-hydraulic parameters such as II ' d< pressure will allow verification of the RELAPS code and model accuracy d in reproducing the test configuration physical geometry and I thermofluid transient phenomena in support of the force calculation validation. i Note that to simulate the test observations, the RELAPS model t f, discharge junction area has been set at 0.03845 ft2 or 15% smaller than the pipe area. The RELAPS-FORCE input listing for this problem is included in Appendix B. 4.1.2 COMPARISONS BETWEEN RELAPS-FORCE CALCULATIONS AND TEST DATA, EDWARDS' PIPE EXPERIMENT Comparisons between the RELAPS calculated thermal-hydraulle parameters t f t-and test data, Figure 4-3 through 4-5, indicates good agreement .k between the experimental and calculated results. Excellent agreement C is also indicated by Figure 4-6, where the measured end thrust load is compared to the RELAPS-FORCE calculated load. I 1 i it ', l \\ l !l m

1 i g... . 1 f f i f f iI Diment.ea l Feet j mm A l 0 26 4 79 8 1 2 74 6 815 aM Henal diameur C e 82 i 555 2 88 h @ mm) l,.

1. U I III Addstional thermecewpies y off)

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( a w O "c 1 c s y w 3 s ,b x e 8 4 s a. a + ,k L n W4 z .4 M w ,N O e Zr g o .f. 3 5 m+6 p =. w o, y-Z s ko3 Z 3

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f i v, i f l m= .a i_ t e ,i C ] s a ,f O m 's) z = 5-r .s, i q yy i i <1 Cl Ci I' ' zl t s t a e ri W L t< m i sa ei u s s a s. sl i. C L, d L I c = . t. c L e m m m c i t =. 2-1 i i t; s i e-i e c r, u e a,., c .J l E i I. i ri e E I s mli d E E ( m, m vi -i es i = a i-z e i i i =5 C L g { m_l u i g j s L != 5i '_1 $i c; I 9 .= u m ) 1 s) e i 4 3, e = s -;i c e, f i t._ c_. i s i 4 d, ei .= c_l u: T' i_ = 3< i s l i A' t. u i i .s e i c, !s ,a e. c g c ccer c cos. o cos c cat co l_ (?NI/JEl) 000090E d i: .s l5' L l4 'E le I ~.. e Imum .) i

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I d- ? I t Mt. [ s 001 O \\ \\ .L v i c \\ C M 2 C \\' \\ N 9 l_ T \\ H = a Cl Of s- \\ p Z, vi \\ A 5 i m i r- \\ U H N \\ E d C t,_ m: H LO \\ I .r C ej 4 I I t C s C Of l -l 9 f O Cl 'g '~ I .O Cs g -f s j h 98 g r e C y 1 U C O, s g ~6 r., 3 g. q. cl L i i. n < r ' '. _ - e!, L.;; .4 v___ s. a, E< O 1

==6 C+ 7 g e, z ~, i t is c El .i a c m m ei s t e c_ P s .s u, >i c, i o ul' i = T i I i '2 Ql g a ci e t, e e s s a t_i c l s e s' 21 a c t f Ti O! i C_ l' Ui g =: 1 ui i ..Fl. c::: i s s = s in e. t i 'g c's e'c ec sc e.o a.{ ~ 00C090c. *-.0A i 't ~ is i..= e NW

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=

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1. 4' s

4 tn C 1 \\c c_ u l C x a s = = _ - _ _ _._ _ _._ ;: t ;, U a x i.______. - ,r - + _ _ i_ i __.________~__I_~.-_-._

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=. 4; t g sr Et r 5 co oo cc. WnS I0Z'1101 I09 IIEC.:! a i ) ~ i 1 4 I.B

~ 4.2 HANSON'S PIPE EXPERIMENT Another excellent sout'ce of test data of blowdown fluid thrust forces is contained in a 1970 report by Hanson(6) The experimental configuration of this pipe experiment is sf " on Figure 4-7. Although experimental pressure measuremem - f, PS1 to PS4, this data was not available in Reference 6. 1' i This " pipe break" e::periment was initiated by over-r ressurizing a rupture disc at the exit a 1.1" ID pipe with a positive displacement pump. Tests were conducted with and without exit oril' ice plates of 30% and 10% of the exit pipe area. 4.2.1 RELAPS-FORCE MODEL OF THE HANSON's PIPE EXPERIMENT The RELAPS-FORCE model of the e:<perimental arrangement is shown on Figure 4-8. The initial water temperature was assumed as 800 F vs. the test's 600 F to avoid early abortion of the RELAP5 calculation due t: a " water-property-error." This assumption should not alter the alts significantly since both the density and sonic speed for both mediums are essentially the same. The reported rupture dise opening time of 0.35 msee was modeled with a linear characteristic opening of a RELAP5 " motor valve" (MTRVLV). Similar to the Edward's experiment, the verification approach will consist of a comparison between the test data and the RELAPS-FORCE calculations. However, only load cell data was available for these I me

3 r comparisons. Thb RELAPS-FORCE input listing for this problem is (' t included in Appendix B. I 4.2.2 COMPARISON BETWEEN RELAPS-FORCE CALCULATIONS AND TEST DATA, HANSON'S PIPE EXPERIMENT Figures 4-9 through 4-11 show the RELAP5-FORCE calculated forces and ) the test measured data for the 3 different pipe exit area experiments. The agreement between the experimental and calculated transient forces is very good with respect to magnitude with small deviations in timing i due to slightly high RELAPS calculated sonic velocities. ( N l t i h L l=i tL m L

7--- 7... ) ~ f q f 1 I l i 40 0 X I_. Vertical and Load N f M Laterol Support cell 1 Orlflee Plate i " 1 1 Pt 3 .l P(l A P4 P2 O h l' - ID 2.323" 1D 1.10 0" ,' f ' m x. LcNoTH Or SM ALL PIPc, NCwc8 WITH NO CXTENS80N 31 $NCHES I_ WsTH EXTENSION A 24=l/1 1 N C H E S WlTM CMTENStDN 9-Se $NCHCS Pe PRESSURC TRANSOUCER ] P2 IS l= 3/ d INCHES FROM RUPTURC DISC Pd 181p=t/ 4 lNCHES FROM AREA CHANGC WHCH NO PIPC EXTCNS6DN WAS PRESENT, THERC WAS NO PRCPSURE TRANSOUCCR Pd. ORIF8CE Dt AMETEMS OF 39/ 6A AND 19/ 31 INCH ARE USED3 AREAS ARE 30 AND 10* OF SM ALL PlPE. THICMNCSS OF ORIFICC PL.ATC (S 2/16 lNCH. ] L Onor:Cc PLATc is RcMovco rOR STuoicS wiTH rULL-CPcN B R e4.c. L

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C R C N O I 1 L J C I 1 G C I u I P T 00 N 4 1 N r. 1 s r N \\ ~. O P._ Te t s l I I : I g I is X_ / .~ I w m S E R r E s E i V P_ l c I_ t. i E P_ 2 o C S i R O t i. N F s r O. t S s S N.. e A P A l L I. f I. C i e R n c r u l t. t a I I r oh8 985 a =8o4 oa 1 osN? e i d f .)y 4 ~ i i OIs 1_IQ t 4OG <l 1 tfOg - i g u gv% i J:u i ias i, is ~ i y I f I i .i i

a u y, 0 1 0 0 e a n mes 3 es + 8 m_ 9 1 4 y 0 's / V 2 O 0 a. 0 N / 0 i, n 7 0 y 2 a 0 a, m C aA es e F N 4 a A r D n I 1 m1 m_ e s l s S a e E / 9 m T F m1 ,c m r s A S s s u R I \\ C O S \\ G c \\ S O 0 p 0 3_ S G i' H i 0 I g J ) ,r T R E s R C E w E s S s s D M u ( L a_

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f I E g E I . G M c e I C P I v g T a 4 t 0 0 r_ 4 3 0 \\ 2 J 3 0 n 1 a N P K.t O X / e v E l I f S r 'I C P i i \\ t n t i g, I t g I 1 t 1 o JY R E I E P w I D V I n t C E e, r U C S 2 e-0 R e O N 0 r, I. o f J F O 0 v r t S s l S N e e P 0 [ O 1 i J r 1 8 I L l i l E t s J3 R e 4 w 0 D. t 00 n l f 7 { n, i .0 2 i

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- 1 i = I = ,==. il s i I ' en 1. I c i Q j v I I-cc s .e N 'e Ol C i ~, i d z. R \\ ~ 1 ts e f .. l N hL f {' wi t ~ a Z 4 4R l B en i N e ui n s L_ i t1 C ', Ln ._ = 3, cc s' d . i a: I i a wl i (n c' ,J Cl, m c; l I eI !g c' I .s a .m.- cr; o t' U e 'n y C C =- .a. e C ~~) u Cl ~ ~. t i o I! c_ H ,1 a j / %"" l

==I t N -t, i. .r ' s = ( z' c_i e C.' xl = g 1 ug p tnn l[. = ? c! U: ~- s u; c_i g e .' 4 c., e yj /% i Ci T: a g ( i cr, -, i u. ~ t s o; = s ? L,. z' = - si i a e_. c_.,. b. t ~ J: s, U-c t i 2; vl 3 s s. s u ~ if 5 R. 's. d sI i'

n O*ObOS O'Ob06 0*0002 0*O C*CCO$-

= i' c.._ 10_4_0_7 1 0 9 _ m_ _':0 _J e 'R (:19*)) .N ...= 6[ e .m

I. 1

5.0 CONCLUSION

~ The comparisons beween RELAPS-FORCE calculated loads and test data discussed in Sections 3 and 4. indicate that RELAPS-FORCE can previde I-good engineering estimates of hydrodynamic piping forces produced by fluid transients. It should be indicated that RELAPS-FORCE simply performs the required computations to calculate the hydrodynamic force 1 histories from RELAPS output parameters and thus the accuracy of the e resulting forcing functions is greatly dependent on the RELAP5 modeling experience of the analyst. = i L l A. L s L s s I e e W ]

\\ l

6.0 REFERENCES

1. RELAP-FORCE User's Guide, November 1983, University Computing Co.

2. Ransom, V. H. et. al., RELAPS/ MOD 1 Code Manual, Vols. 1 and 2, EG&G Idaho, NUREG/CR-1826, March 1981.

1

3. EPRI/C-E PWR Safety and Relief' Valve Test Program, Summary Program Description, EPRI, 1980.

I

4. A. J. Wheeler,and E. A. Siegel, " Measurements of Piping Forces in a l

) Safety Valve Discharge Line", ASME Paper No. 82-WA/NE-8. ] i

5. A. R. Edwards and T.

P'. O'Brien, " Studies of Phenomena Connected i with the Depressurization of Water Reactors," Journal of the British f I Energy Society, April 1970, vol. 9, pp. 125-135. Lj j F, l~i

6. G. H. Hanson, "Subcooled. Blowdown Forces on Reactor-System j

Components: Calculational Method and Experimental Confirmation," IN-1354, June 1970. 1 i:. ~ me l k

n s. i I, I v 1 b, \\' l 4FPE>J Vm A 1 g usun,ME4Tf, OF E l? t4f, FOccEt IN A SATI.Ty VAL /t. St$c\\ACat M Lt Q 1. i ~ L' 1 1 6 L I % d k w

.v ur s THE AMERICAN COCIETY OF MECHANICAL ENilNEERs 82-WA/NE-8 346 EA7 St., New York. N.Y.10017 El + V .3. ne ssee.v en.n,mi.e. -. e.emems., - a ene .n.ees.o.,m u f c.

• -- er mentmos ei sne sec.eev er et no oweene er sectone, or onmaa m no i

I g - 2. Discussen,a onnteo enry it tne peser a oueinereo m an AsMs soumes.. Roosened for penefet DuclacetaOn upon pressfuetMm. Full Creelt ermutd De green to ASME. ene Tecnnicas omeen, ena one autnenn passes are evennem hem AsME ter nme emnene 4ttertne msetsng. j Pnneed m USA, MEASUREMENTS OF PIPING FORCES IN A SAFETY VALVE DISCHARGE LINE A.J. Wheeler Project Manager Electric Power Research Institute Palo Alto, California ) l E.A. Siegel-. Principal Nuclear Engineer i. Combustion Engineering. Inc. Windsor, Connecticut I 1 ABSTRACT nected to the pressurizer such that the

• inlet to the valve is' exposed only to steam. In another comon Measurements were made of support reactions to design, the portion of the pipe inmediately upstream I

transient hydrodynamic forces on the discharge line of of the valve contains liquid water. This design, ) a nuclear reactor safety valve test facility. Data commonly called a loop seal design (Figure 2), is used i is presented for three different test conditions - to minimize leakage through the valve seat. When the two with upstream loop seals and one with only steam. valve on a loop seal design opens, the water is pro-Sufficient information is provided to permit verifi-pelled down the discharge line ahead of the steam. i I cation / development of hydrodynamic force predictive t-

models, Due to its high density, the water produces signifi-cantly higher loans than steam alone. The EPRI pro-INTRODUCTION gram obtained piping load data from configurations both with and without loop seals. This paper prasents 1

For worker safety and operational reasons, the data for three tests - an all steam test and two loop seal tests, steam discharge of nuclear reactor safety or relief valves is normally routed througe pipes. to a tank of C water which quenches the steam. When the valves ~ open, SQ )$ ,m the acceleration of the fluid in the discharge pipe causes substantial transient loadings which must be restrained with suitable piping supports. Estimates fg of the transient loadings are required in order to design these supports. me L Although analytical me'thods to predict these loads have been available for some time, the data a base to benctanark the methods has not been extensive, 1 I and significant uncertainty has existed in predicted [ loads. As part of a recent safety and relief valve test program for pressurized water reactors (PWR's) managed by the Electric Power Research Institute (EPRI) and sponsored by the utility owners of pres. i .( I surized water reactors, an experimental discharge- \\/ i' line was instrumented with load cells at the supports {8 8WW to extend the data base. This paper cocuments key tests performed with the Crosby 6M6 safety valve, the most common valve used on PWR's. FIGURE 1 - SKETCH OF PWR In the pressurized water reactors design, the PRES $URIZER AND SAFETY VALVE reactor safety and relief valves are connected to the pressurizer, a pressure vessel containing both water and steam which is used to control system pressure (Figure 1). In certain transients, steam is released ~ through the safety or relief valves to limit the sys- ' tem pressures. In some oesigns, the valves are con-

s ,s y l t; t ri f four segments is supported at one end of that segment in the axial direction. A design goal for the test facility was to Ininimize or eliminate load path re-dundancy for all discharge piping loads. This goal has been attained by providing axial pipe suoports whien are very stiff compared to alternate load paths. msms The elimination of load path redundundancy has been 6 "g confirmed by dynamic analyses as part of the overall s 8" -g design effort. [hV Each segment of the pipe was supported with two load cells, arranged in parallel. The net axial load { l on each segment is the sum of two load cell measure-j ments. Data was recorded with a digital data acquisi-tion system which sampled the load cells at a rate of Z 1000 samples /second to allow resolution of up to 200 a HZ. The measurements are accurate to 12200 N. Static pressures were measured at several down. stream locations threr of which are presented here - PT09. PTIO and PT11. These instruments were also { FIGURE 2 - PWR LOOP SEAL DESIGN sampled at 1000 samples /second to allow resolution of g frecuencies up to 200 HZ. The accuracy of these mea-The primary objective of this paper is to make surements is 3 28 kPa; however, the sensing 1.iaes, data available to individuals developing or verify. which are about one meter long and filled with cold 3 ing mocels to predtet piping loads. water, might tend to reduce the accuracy at high fre-r

quencies, g

TEST FACILITY Three other important measurements are valve flow-The test facility (Figure 3) consists of a four rate, valve stem position and upstream reservoir tank p segment discharge line attached to a spring-loaded pressure. The valve flowrate was measured with a g safet valve In all cases presented here, the venturi which is only accurate under steady or close piping upstream of the valve was a loop-seal design; to steady conditions. In the test facility these con-the loop seal water was drained for the all-ditions occur a few seconds after the valve has open. however'sts. ed. This measurement is accurate to about + 6t of steam te reading. The valve stem position was measured direct-ly with a linear velocity and displacement transducer,

men, This instrument, accurate to + 0.13 mm, was also sampl-m'r i

ed at 1000 samples /second. tee upstream vessel pres-sea. Ex N sure presented here is PTS 2 an instrument sampled at 3 I h/, 100 samples /second to allow resolution of frecuencies "'"'"d If 5" *. up to 20 HZ with an accuracy of 6 1 9 kPa. nmas e -y' a es ur For loop seal tests, the temperature of the loop 1 % seal water is significant. If the loop-seal tempera-l T ture is high enough the water will flash as it pas. i '= 'r ses through the valve, reducing the measured down-tr stream loads appreciably. Two thermocouples were "7" utilized to measure loop-seal fluid temperature direct-ly. In addition, several thermocouples were attach. it in su me ed outside the pipe to measure wall temperature. It w a s p h" % is expected that these measurements can be used to de- .I 8 "l' "nU. duce the fluid temperature axial distribution. y wnan .am eru. STRUCTURAL EFFECTS IN THE MEASURED LOADS Allrasu / When any structure is subjected to dynamic loads. '; sr e' sen-it will respond in a manner that may either amplify g

hmm, or diminish the applied dynamic loading. This aspect lE

.r 9n of dynamic response characteristics of a structural 4 - M support system is referred to as transmissibility. y,,"-Q'85 8' Transmissibility is defined as the ratio of the sup. ( port load to the applied load. The magnitude of this 4 ratio is dependent on the relationship of the cyclic h frecuency of the applied load and the natural frecuency of the structure. For an infinitely rigid structure, this ratio is unity. For structures with finite stiff-h FIGURE 3 - TEST FACILITY (1 f t.=0.3048m) nesses, however, this ratio can ce significantly above /* or below unity. The hydraulic loads act in the axial direction in each of the four discharge pipe segments. Each of the 2 L \\, l

>e I i Verification or evaluation of a fluid model for o, predicting hydrooynamic loads uses a process which is similar to the application of the same code for power g i plant design. First, the fluid code is used to predict h au'* ",/ the hydrodynamic loading on the inside of the pipe. Next, the hydrodynamic loadings are used as forcing. m function inputs to a transient structural model of the piping-support system. In the case of plant -) Lf. un i j' design, the resulting support loads are used to design a u]l dia the supports (probably with structural reanalysis). 1 r In a verification effort, the support loads are com-pared to the test data to evaluate model adequacy. g fluid loads directly but rather the response of the I / so.mo e The load cell instruments do not measure the j 6.* - ~/ j j piping-support system structure. Every attempt was = made to construct the facility 50 that the supports l i were very stiff so that the load cell measurements E I were as close as practical to the fluid forces. s Nevertheless, it is necessary to include a model of 9 l the structure to fully validate fluid loading computer [ ) I codes. g In order to design the test facility, a detailed structural model of the whole facility was developed. vm I = i 1 Based upon test data which gave infomation about the l IJ natural frequency characteristics of the facility, the structural moeal was adjusted to represent the as-built facility. However, not only is that model l]l-much more elaborate than is reouired to evaluate discharge piping loads, but a detailed description urnl 5 of the model is beyond the scope of the present paper, w n.ri su _ i As a result, a simplified structural model based on the more complex model has been developed. This FIGURE 4 - PRESSURE VESSEL. SUPPORT SK!RT l model consists of the reservoir pressure vessel, the AND VALVE INLET P! PING 1l valve, the discharge piping and the piping supports. (DIMENSIONS IN INCHES, [ The recuired dimensions for the tank and valve is (1 INCH =25.4 mm) {' shown in Figure 4 Each of the piping supports is tse eru uan s represented by a spring or springs as shown in Figure ca so. .o,c no. .i ) l 5. The second, third and fourth segment supports ur rs '{ L attach directly to the concrete basemat. The first 'y l segment supports are attached to a beam structure g i which in turn is attached to the reservoir pressure n i vessel. For structural mocel purposes, the beam u u d'" omens smus ris' wm l structure mass is approximately 9070.0 kg and located l

go q'

at the top of the pressure vessel. The flexibility r llj s i e.res i @' aen j of the beam structure is incorporated in the stiff-i aae o.n n=sicum ness of springs 1 to 2 ard 3 to 4 in Figure 5. The ~we 8 i vessel in turn is cantilevered from the basemat. It \\L Ms'Q" j is considered necessary to model the pressure vessel 3 Ro emtsuon @,Msuoi, '~ and skirt stiffness ar.d mass in order to provide a 11 se samme j@gj minimum system model. jjgstt ' u u smio H* 14 to Es0La0 t TEST RESULTS a s The results of three representative tests are re- .6 - Y' .u f na / { ported here. These tests are distinguished primarily f f', by the inlet candition to the valve. The tests pre-

[

1 is sented are: [ 4 i suo s3ss s s 3 Steen only - Test 1411 i FIGURE 5 - STIFFNESS OF P! PING SUPPORTS ? Sttam with a cold loop seal - Test 908 (1 lbf/in=0.175 N/m) 4-Sceam with a hot loop seal - Test 917 o 4 l ,i M1 of these tests were perfomed with a valve manufactured by Crosby Valve and Gage Company, model e HB.P9 86 It has six inch (152 m) inlet and discharge ~ m}s Mes. The valve has a nozzle area of 2.35 x 10'3 f '~ ano an ASME rated flow of 191,615 kg/hr with 17750 h>a saturated steam. ~ l 3 ~ l

L Test 1411 Figures 8 through.11 present the measured forces This test was performed with the loop seal por. on each of the four pipe segments. The peak forces h' tion of the inlet piping drained of all 11ould so the observed on segments 1 through 4 are 11565 N, 28912 N. valve was exposed to a steam only inlet condition. 53,376 N and 66,720 N respectively. The peak forces J Prior to the test, the valve was leaking slightly, on segments 3 and 4 are somewhat delayed. These t Thig appeared to heat the downstream pipe walls to peaks are thought to be caused by condensate (result. (l ing from pre test steam leakage) striking the elbow 100 C and replace the air with steam. The quality of the downstream steam is unknown, just upstream of the pipe discharge to atmosphere. The pressure history upstream of the safety valve is shown in Figure 6. The safety valve popped open gI .Z(..j!..~l. l. 1 when the tank pressure reached 2410 psia. The valve s- '-j.. stem position history is shown in Figure 7. After the 5 initial transient had subsided, the cuasi. steady flow l, g T-through the valve was 213,145 kg/hr wnen the tant gl } pressure was 18234 kPa. } l gg j j fl f. i fg ' n.- - W.,..-..w., ).. a.. g % q - %...a p E j.- g l b [ 5 j jj - 2 g. i . ~.]. g 'i o -j g l g; -+ ,B,, t y I t \\ 23.25 23.64 n.4 2L 92 g N j sfCoMos 'l~~~".~.~... 5 l FIGURE 8. TEST 1411 SEGMENT 1 FORCE HISTORY ~'..g. f j (I lbf=4.448 N) b jp.Q M,_.:7~ 20.o 23,3 3a.0 44.0 st:ms FIGURE 6. UPSTREAM RESERYOIR PRESSURE HISTORY ( (1 psia =6.894 kPa) { j l 's, .... _._ I.

5....

g s ___._...q... ,.3.. ,---- -4-~ {- ' '} ~ ( "5- "- N- [... I ,.yi.. _, _ = = i, 721-1.:. ;__..JI l -Ihr 1 j j t_=1 --d_ b f I j !s . _ __; _ ;.- - +-- l i j ) ._... r...._ 4 -(=1. : E !-J l N. !.Q.t.....,,,.i4I = o-s: g -~ ~= n. g 1 t i 9 -.: ( -.M * : l~. i f

. rl.=__

-U;:'.l' - L. ; l. : : !_......'....._../. i] i l 0 Id I .~ s - ~: I e n.2s n.ss n.n n.n L n.ts n.22 . n.<a n.sa acuos neanos FIGURE 7 1TST 1411 VALVE STEM POSITION, ZE36 FIGURE 9. TEST 1411 SEGMENT 2 FORCE HISTORY (1 inch =25.4 m) (1 lbf=4.448 N) u 4 \\ !m k

s I I l i 1 i af,

• • l

l l I i t 5 i l !A T-o J l [-

g 4

~...;.... ...t. -. j i g 1.. ~ .i - -. l4 i. g, -~ {, f. g ~ f._ ';fl[ i 4 --1--. g l-I g i t Ji l y-1. .} .l = I, t ' v I' r 0 j g 3 I I 6 M I I i i g i j 23.28 23.44 23.60 23.76 l g steencs I 23.2s 23.aa 23.so 23.76 FIGURE 12 - TEST 1411 STATIC PRESSURE TAP i sican PT09 RESPONSE I FIGURE 10. TEST 1411 SEGMENT 3 FORCE HISTORY (1 lbf=4.448 N) .R. i I i -a. 1) = g 'l 2 8

f.i i

= = l i s . _ -.' t,~_..r-h } i p - Q, d-- __, _.g 4 B= 4. _ j_ _._ i y i E* ( a

== ._.i-_; ._i_. ._4,_ y i l i g - -y }J.. = i j 3 - ~. y..- ! g ... -; -~. 4.. j... ;, g..., i j E* . _i _... i. ___ ; _.. ] ~ i ,} j.. 4. .f}, E 23.28 23.44 23.6o 23.92 3 g ... p._ p...- P - ' h 1 -t* stCCNos i 8 _ +. _ ! q .. ~ 9-.: -q._... .j-- r FIGURE 13 - TEST 1411 STATIC PRESSURE TAP j 0 - - - -*~ - - ' - - l' PT10 RESPONSE i i. 23.2s 23.44 23.60 23.7s (1 psia =6.894 kPa) j stCONDs 2 FIGURE 11 - TEST 1411 SEGMENT 4 FOR".E HISTORY i } l l l 1 i l' I (1 lbf=4.448 N) 5 i g f { [c Although not a measure of force, static pressure i I data is also useful to validate analytical models of i= pining hydrodynamics. Pressure data taker at taps PT09. PT10 and PT11 are presented in Figures 12 { hg j througn 14 Ten seconds after the valve popped open r 16994 kPa. and these three pressure taps gave read. Si f l{ {fr and the flow was ouasi. steady, the tank pressure was i 4 q ings of 1551, 733 and 738 kPa respectively (not shown g3 .q on the Figures). i r I Test 960 N: l Inis test was performed with the loop seal pip. E

1. NI e

i i ing filled to the top with cold water. The tempera. j ture measurements are presented in Figure 15. For 23.2s 23.44 23.so 23.7s this test (but not for 917 or 1411) a 99.3 mm diameter s!conos orifice was installed in the 12" pipe just upstream y of the discharge *to atmosphere. Tka downstream pip. ing was at 15 38 C prior to the test. FIGURE 14. TEST 1411 STATIC PRESSURE TAP PT11 RESPONSE (1 psta=6.894 kPa) L. 5 J

.= 1 r m (i s i ~ ~ = " ' - - . J'" 4 { f_ n _.

• fl%

g i.

• r~m

p, 1

m. m.- - g*. = e a e m m. m.- _. - y. :-. -g. m. m.m m; im .m. m.n 9 a: - =. ij i

E:

' p.y_ t~""- 1 . m.n ,_ i / w e m an - +--- 1 m. m.n j )\\ j w s-~ i.e. - EF j ( p , z i-FIGURE 15 - TEST 908 TEMPERATURE FROFILE h ' * ~~ t ' ~ IN LOOP SEAL t-~~~ -~ j .4...."~. ... ~ ~i1;p! g (tc=(tf.32/1.8) Ej ____i The tank pressure history is presented in 32.00 12.so 33.so 9.4o Figure 16. The valve-stem position data (Figure 17) indicates that the valve did not simply pop open as in-sacanes. 7 test 1411 but instead oscillated for about 0.6 seconds. FIGURE 17.- TEST 908 VALVE STEM POSITION ZE17 i remained partially open for about 0.3 seconds and then (1 inch =25.4 mm) popped fully open. This type of behavior was observed .) on most of the spring-loaded safety valve loop seal tests run during the EPRI program. The flowrate was

• ).

observed to be 210,900 kg/hr wnen the tank pressure =2 i was 18545 kPa psia. In this and test 917, the flow. g . - p. \\ a i rate is the steam flow in the quast' steady condition 5 e. ~ t-( 0 i Ai 4 after the loop seal had cleared the valve. r ..N I \\) ) M. N. -e

r

- N g -- . li *l

• i

\\* l n g i a g

• - A a

l l g I j . \\g [ SE g

4. -

u.t s u.23

x. 31 m.3s j

stconos El I _,i FIGURE 18 TEST 908 $EGMENT 1 FORCE HISTORY = i i (1 lbf =4.448 ti) 1 I i 1 i l e i 30.0 3a.c 46.0 54.0 'e' -i'----+.. ^ * ~ ~"4 ~'~~ ".'~g -.f "' * ' l - *

• sM

'!I -- l - - FIGURE 16. TEST 908 UPSTREAM RESERVOIR k ./. f i ! ) -- -A--lV p! k @ PRESSURE HISTORY a u (1 psia =6.894 kPa) !2 il I The observed f"ces on each segment are shown y* hV... l. 1 ~.-.- l. ~.- i... j l ' in Ffgures 18 - 21. The observed peak forces on seg. - f - - i- ~ I-4 ments 1 through 4 are 97,856 N,800,640 N. 266,880 N h ..q. y f. 'l I and 88.960 N. The large 800.640 N load appears to be . + - -.. l. i -~ caused by a relatively intact water slug striking the S

• • 4-~+'

i-i second elbow, ho such large load is observed at the

g 4

j .9 ,a third elbow - the slug may have partially dispersed 1 i by the time it reaches that point. On the other hand, E i i a large load at elbow three may be masked by a combi- .i nation of the structural softness of the segment 3 .A --i -i -, i i supports and the discharge orifice and short length 3* t* 'i ~ i of segment 4 M.1s u.23 x.si

u. 3s stconos FIGURE 19 - TEST 908 SEGMENT 2 FORGE HISTORY

~ (1 lbf =4.a48 N) 6 \\ i l 1 L

.v .s. a i ~. f 4" r i o j. r E p._ p___p 8 a ---i- - - - - _i i ~~ s -q= J, j : c e -+ I t = a., -.--... - f. ~, 8 1 --'--~f---- y*- Q, s l = y g. } x 1 g y / i A-1

y. v;.:.

,. -.s V -l .a. - u _ 4, 8. g "', i 4.15 M.23 M. 31 34.39 5 steam l M.15 M.2' N. 31 M.39 ' FIGURE 22 - TEST 908 PRESSURE TAP PT09 RESPONSE. g stems (1 psia =6.894 kPa) I FIGURE 20 - TEST 908 SEGMENT 3 FORCE HISTORY (1 lbf=4.448 N) $g ._. i _. 1 P. r. 4 [ g p_. ...l._...ip. _. _. y.g __u o s 1-a- s

}~-

= - - - + - ,3 E-

5. -

p..._ }. $

... { . j.. _ t-s. _ 4_. _ _ _. c 8 - fi.. E p....ji. L 4., .- j s .. j., a.H u .5 L l ..[ _J .- @ -_l -1 (. ' g o, t g 'i -.i k' -M ~ ! "d E M.15 4 M.23 ' M. 31 M.39 l .'.".. h:'_ ~ T~~'I. . y.h g stems } E _.4 y i --4---- ---i- - E. FIGURE 23 - TEST 908 PRESSURE TAP PT10 RESPONSE E =.. _ l.- e --t (1 psia =6.894 kPa) =. j .. j . y... p. ._.._.~f.. j-M.16 M.23 M.31 M.39 E t acm= a 3 _1._.._. ~~~I"~~~ FIGURE 21 - TEST 908 SEGMENT 4 FORCE H* STORY t 'I (1 lbf =4.448 N) 3, The static pressure measurements for PT09. PT10 ha and PTil are shown in Figures 22 - 24 Not shown on g l s. i. the figures, wnen the system had been flowing for = 3 i about 10 seconds and the pressures are quasi-steady, w _wj j i y these pressures reached values of 4481., 4481. and l l . g g 4550. kPa wnen the tank pressure was 16,959 kPa. -l 4 2 i

• M.11 M.23 M.31 M.39

_ Test 917 ucum inis test was similar to test 908 except that the loop seal volume was filled with fairly hot (1770C) water prior to the test. It was expected that this FIGURE 24. TEST 908 PRESSURE TAP PTil RESPONSE wuld reduce discharge piping loads since flashing of (1 psia =6.89 kPa) the licuid would disperse the downstream water slug. q The temperature measurements on this slug are shown g in Figure 25. The discharge piping was at amotent Figure 26 shows the tank pressure history. temperature (10 380C) prior to this test. The valve Figure 27, showing the stem position, indicates that flowrate was observed to be 204,075 kg/hr when the the valve oscillated much the same as in test 908 tank pressure was 18,821 kPa. t I g A

1..

7 1

j I The segment loads for this test are presented in ,f, Figures 28 31 The observed peak loads are 31,136 N, 66.720 N. 77.840 and 65.386 N sn segments I through

4. respectively. These represent significant reouc.

tions in loading relative to test 908, particularly m e e e m' w. m.re on segment 2. l ,, I I, w. m.n L8' t w. ses.sg = m 9 w a san.n 2 w N.n ~ Iet

• *** h a

.l W, t I

• 8'"e d

,g e+ i. i tem. ar in. ' o! \\, [r. . i

s y

,1 -- j 1 FIGURE 25 - TEST 917 LOOP SEAL TEMPERATURE PROFILE g f

' ; l'p, l) j.

(t - (t.32)/1.8) r = i

f. k ' f I

k 3 l ad n E dg f' l ~ =

• 18.10 18.20 14.30 18.40 g

,1 stCONDs g' - ' f a 4 * 'i i FIGURE 28 - TEST 917 SEGMENT 1 LOAD HISTORY .. j ' ' ' ' ' l l' (1 lbf=4.448 N) xs . lii i I i i ) w ' _ ). i~ q,;.7;y ; ; 7. ja. g j j .f... p. _. y. r9 A i s-... + ...........i. S f - I 4 . 4 w. ._..j. p. = j'. __ g. .i_..... 18.0 22.0 32.0 42.0 ' h. g.._. y l

p*-

~t.. \\ sim0s u. I at ._.t 4 jl FIGURE 26 - TEST 917 UPSTREAM RESERVOIR 5 [.. j._., g. PRESSURE HISTORY s A I-(1 psia =6.894 kPa) l- - i.. I W. i ._..3. il ..g.. .l. ... -. j... ;..,., i =7 ..... ~ ?b.~ ?. T. i .4 i g .~.7 _ a j,: 18.10 18.20 18.30 18.40 j Tj srconos p. i l.: .) d 2,,,,

• d P-FIGURE 29 - TEST 917 SEGMENT 2 LOAD HISTORY 2: *

(1 lbf=4.448 N) 9 7

• g

-- ! =e = The pressures at PT09 and PT10 are shown in 3 i Figures 32 and 33. Ten seconds after the start of ~ E J i the transient, these pressures had values of 1585 . ~ ii ! and 745 kPa when the tank pressure was 17.510 kPa. E.' "'"n=="'" These cuasi-steady pressure measurements are not shown d i j on the figures. 45j "--'~~~

.b i[h g

16.a0 17.60 18.40 3 srcanos y 1 FIGURE 27 - TEST 917 VALVE STEM POSITION HISTORY l ( (1 in=25.4 m) 8 \\ \\ l I i

.j % 3 g ~ ~~ 4 ~. ^ s: MSpWua 2 II ~ r 'l ^i in L'~~ L. f. i V i. / s. i

. j...

'l = i k 4 2 4-T U i .i. l 5 'J.Q. ' " J' $2 ' f' II F

i j

j i y I i 18.10 18.20 18.30 14.40 i srcons '= l e i 18.1o 18.2o 18.30 18.40 FIGURE 30 - TEST 917 SEGMENT 3 FORCE HISTORY sitons (1 lbf=4.448 N) l FIGURE 33 - TEST 917 STATIC PRESSURE TAP PT10 RESPONSE 1 (1 psia =6.894 kPa) l E 9 CONCLUSIONS O e. WW1\\AA k The following conclusions were reached in this u t ef fort : i g y M .j-1. These test results provide a data base which 5 g s can be used to benenmark fluid codes to i I y g. predict piping hydraulic loads. _ p _ w. k H h. g 2. Cold loop seal designs can have substantially j, higher loads than steam only designs. Je V T ~-" 3. Heating

• loop-seal liquid into the range i

( 18.10 18.20 18.30 18.4o 300-350 F can substantially reduce peak srcam piping loads, f FIGURE 31 - TEST 917 SEGMENT 4 FORCE HISTORY 4 Accumulated condensate (from leaking valves) Ll. (1 lbf=4.448 R) in inadeguately downward sloping discharge lines may produce nigner loads than would be expected from a steam only discharge analysis. E" ACKNOWLEDGEMENTS 5 I E 3 t' The authors wuld like to thank the staff of J 3 Combustion Engineering. Windsor, Connecticut, who i l B ^ l designed and operated the test facility. Special thanks go to Mr. S. Austin for nis efforts to modify the system structural model. .~2 I l. The authors also thank members of the EPRI staff { who participated in the preparation of this paper. g i f " e' i 18.10 18.20 18.30 18.40 stconE FIGURE 32 - TEST 917 STATIC PRESSURE TAP PT09 RESPONSE (1 psia =6.894 kPa) 4 ( 9 6

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