ML20149D771
| ML20149D771 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 01/06/1988 |
| From: | Tiernan J BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| TASK-2.D.1, TASK-TM NUDOCS 8801130118 | |
| Download: ML20149D771 (15) | |
Text
B ALTIMORE GAS AND ELECTRIC CHARLES CENTER R O. BOX 1475 BALTIMORE, MARYLAND 21203 JostpH A.TIERNAN Vict PatsiotNT NUCLEAR ENenOY January 6,1988 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:
Document Control Desk SUBJ ECT:
Calvert Cliffs Nuclear Power Plant Unit Nos. I & 2; Docket Nos. 50-317 & 50-318 Thil Action item II.D 1, Performance Testing of Relief and Safety Valves R EFERENCE:
(a) Letter from hir. S. A. hicNeil (NRC), to hir. J. A. Tiernan (DG&E),
dated April 30, 1987, Request for Additional Information Gentlemen:
On April 30, 1987, you asked for additional information regarding our implementation of Thti Action item II.D.1, Performance Testing of Relief and Safety Valves (Reference a).
The attachments to this letter contain our response.
Shou!d you have further questions regarding this subject, we will be pleased to discuss 1
them with you.
Very truly yours,
-T M
i i
l i
JAT/WPht/dlm Attachments eu D. A. Brune, Esquire J. E.
Silberg, Esquire R. A.Capra, NRC d]
S. A.htcNeil,NRC W. T. Russell, NRC T. Foley/D. C. Trimble, NRC O
\\
in"RasBE$J7 P
ATTACilMENT A RESPONSE FOR REQUEST FOR ADDITIONAL INFORMATION TMI ACTION ITEM II.D.1 PERFORMANCE TESTING OF RELIEF ANDSAFETY VALVES NRC's questions is restated, followed by BG&E's response. Background information that was included in NRC's more lengthy questions was removed for brevity.
l.
The plant block valve is Velan B9-354-B-MS while the test valve was a Velan B10-3054B-13MS. Discuss the differences in the valves due to one being an MS and the other being a 13MS. Discuss what impact these differences may have on valve -
operability.
RESPONSE
In previous correspondence we inadvertently referred to our block valve as an MS.
In fact, they are Velan B9-354B-13MS We regret the error.
2.
Provide the torque produced by the Limitorque SMB-00-5 operators. If the torque is less than 82 ft-lbs (the minimum torque tested by EPRI), provide test data to demonstrate the operators are capable of providing adequate torque to close the block valves.
RESPONSE-The operators will be tested using MOVATS during Unit One's spring 1988 refueling outage. The results will be provided to you upon completion.
3.
Provide the maxin'um expected backpressure and bending moment for the Calvert Cliffs I and 2 PORVs.
RESPONSE
Calculated steady-state backpressure assuming two PORVs and 2 SVs open:
Unit 1 Unit 2 833 psia 10%
653 psia i 10%
Calculated bending momentr Valve Unit i Unit 2 ERV-404 1214 ft-lbs 1495 ft-lbs ERV-402 1460 ft-lbs 2380 ft-lbs ATTACilMENT A RESPONSE FOR REQUEST FOR ADDITIONALINFORMATION TMI ACTION ITEM II.D.I PERFORMANCE TESTING OF RELIEF AND SAFETY VALVES Bending moments include loads due to dead weight, thermal, seismic anchor movement, seismic, PORY discharge, and SV discharge loads.
4.
Verify that DC&E has installed the heavier springs recommended by Dresser for PORY operation at less than 100 psig.
RESPONSE
BG&E has installed heavier springs consistent with Dresser's recommendation.
5.
Provide documentation to show the PORY control circuitry has been qualified under 10 CFR 50.49. Alternatively, provide information to demonstrate the control circuitry is qualified under NUREG-0737.
RESPONSE
Negotiations between the NRC Staff and the Nuclear Utility Group on Equipment Qualification (NUGEQ) resulted in the reformulation of this question. Reprinted below is the newly worded question as taken from a July 8, 1987 request for additional information regarding San Onofre Nuclear Generating Station, Unit 1.
Our response is directed toward this question and not the one stated above.
NUREG-0737, item II.D. ! requires that the plant-specific 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 Nuclear Regulatory Commission staff has agreed that meeting the licensing requirements of 10 CFR 50.49 for this circuitry is satisfactory and that specific testing per NUREG-0737 requirement is not required.
Therefore, verify whether the PORY control circuitry has been reviewed and accepted under the requirements of 10 CFR 50.49.
If the PORY circuitry has not been qualified to the requirements of 10 CFR 50.49, provide information to demonstrate that the control circuitry is qualified per the guicance provided in Reg. Guide 1.89, Revision 1. Appendix. E.
As an alternative, the staff has determined that the requirements of NUREG-0737 regarding the qualification of the PORY control s
circuitry may be satisfied if one or more of the following conditions is met.
A'ITACHMENT A RESPONSE FOR REQUEST FOR ADDITIONAL INFORM ATION TMI ACriON ITEM II.D.I PERFORMANCE TESTING OF RELIEF AND SAFETY YALVF3 a.
The PORVs are not required to perform a safety function to mitigate the effects of any design basis event in the harsh environment, and failure in the harsh environment will not adversely impact safety functions or mislead the operator (PORVs will not experience any spurious actuations and, if emergency operating procedures do not specifically prohibit use of PORVs in accident mitigation, it must be ascertained that PORVs can be closed under harsh environment conditions.).
b.
The PORVs are required to perform a safety function to mitigate the effects of a specific
- event, but are not subjected to a harsh environment as a result of that even:
c.
The PORVs perform their function before being exposed to the harsh environment, and the adequacy of the time margin provided is justified; subsequent failure of the PORVs as a result of the harsh environment will not degrade other safety functions or mislead the operator (PORVs will not experience any spurious actuations and, if emergency operating procedures do not specifically prohibit use of PORVs in accident mitigation, it must be ascertained that PORVs can be closed i
under harsh environment conditions).
d.
The safety function can be accomplished by some other designated equipment that has been adequately qualified and satisfies the single-failure criterion.
Our PORY control circuitry has not been qualified to the requirements of 10 CFR 50.49. However, the PORVs are not required to perform a safety function to mitigate the effects of our design basis events.
It is expected that the PORVs will open during any event in which the RCS pressure exceeds the PORY setpoint. Should the PORVs fail to close, operators will be alerted via acoustic flow monitors downstream of the valves. Digital indicators are provided on control panels C06 and C31. Additional indications are provided through the plant annunciator. Operators will then close the PORY block valves located upstream of the PORVs, terminating flow.
j All electrical equipment for the acoustic monitors and PORY block valves are qualified in accordance with the requirements of 10 CFR 50.49 as applicable.
3
A*TTACHMENT A RESPONSE FOR REQUEST FOR ADDITIONAL INFGi(M ATION TMI ACTION ITEM II.D.1 PERFORMANCE TESTING OF RELIEF ANDS 4FETY YALVES 6.
Your response to Question 2 of our initial RAI stated the liquid uischarge case for the PORV during low temperature overpressure protection was not analyzed because it was not considered a design basis event for the piping analysis.
Also, it was stated such transients occur only after the plant is in a safe shutdown mode and, therefore, do not constitute a safety concern. This is not an acceptable response.
- First, the fact the plant is undergoing a
transient indicates the plant is not safely shutdown. Also, the intent of NUREG-0737, item II.D.1 was to show the overpressure protection system, including the piping, will be able to handle all loads imposed by overpressure transients. It is not acceptable to say, as was implied by BG&E's response, that because the plant is in a safe shutdown mode, damage to the PORV piping would be acceptable. Provide the results of a
structural analysis using conditions typical of cold overpressure protection at Calvert Cliffs 1 and 2 for our review. Include a comparison of calculated and allowable stresses for the most highly loaded locations. Alternately, provide data to show loads during a low temperature overpressure transient are bounded by other tran lents and accidents analyzed in the FSAR.
RESPONSE
We disagree with your characterization of the intent of NUREG-0737, item II.D.i and believe the conceru described in your question is outside the scope of this TMI item. The "Position" statement of item II.D.1 in NUREG-0737 states:
Pressurized-water reactor and boiling-water reactor licensees and applicants shall conduct testing to qualify the reactor coolant system relief and safety valves under expected operating conditions for design-basis transients and accidents. (emphasis added)
Since a liquid discharge event during low temperature operation is not a design basis event at Calvert Cliffs, it need not be considered in response to item II.D. l.
However, we wiieve your concern may be a valid one with respect to the review of our low temperature overpressure protection system that resulted in your safety evaluation dated November 17, 1977. As such, we will continue to investigate this concern and will report our findings to you. In a follow-up conversation w".th your staff, we learned that it may be possible to crimp piping or elbows in the PORY lines, thereby restricting flow during a low temperature overpressure (LTOP) event. In the conversation, the NRC Staff alluded to analyses that have been performed that demonstrate this crimping, it would be most effective if we could apply those analyses to Calvert Cliffs to begin to address your concern.
Therefore, please provide us with the details of the analyses you alluded to in our meeting.
De assured, we will be pursuing the LTOP concern with the NRC staff but it should not be a condition to completing this TMI item.
L '
l
r ATTACIIMENT A RESPONSE FOR REQUEST FOR ADDITIONAL INFOF I ATION TMI ACTION ITEM II.D.I PERFOR.MANCE TESTING OF RELIEF ANDSAFETY YALYES i
I 7.
Provide information on the ve ification of REPIPE. Provide comparisons of the results for REPIPE calculations and EPRI/CE data to verify this code is an appropriate tool to evaluate p; ping di. charge transients.
RESPONSE
The REPIPE computer program was used as the force post-processor to RELAPS for the Calvert Cliffs, Units I and 2, Pressurizer Safety and Relief Valve Piping Transient Evaluation. This program was developed by Control Data Corporation, and has been validated by Pechtel Power Corporation for nuclear power plant applicatians. REPIPE solves the one-dimensional momentum equation for force on a i
control volume (Ref.1) using time-dependent thermodynamic parameters output from RELAP5. Individual control volume forces can be assembled to obtain net reaction t
forces on pipe segments, typically bounded by elbows or large reservoirs (for pipe rupture consideration). The REPIPE program meets all of the requirements of the Bechtel Power Corporation Quality Assurance Manual and Engineering Department Procedures regarding "Standard Computer Programs." A complete set of user and l
?
validation d numentation, including recommended modeling guidelines, is maintained fo the program. The documentation is considered proprietary, but can be made avai sble for audit by the Nuclear Regulatory Commission at any Bechtei l
office on request. The Validation Manual addresses typical applications of the l
REPIPE program when used in conjunction with the thermal-hydraulic analysis prograra RELAP5. Industry standard benchmark problems are used for comparison to e
program results. The benchmarks include Edward's and O'Brien's experiments (Ref.
- 2) and lianson's subcooled blowdown force experiments (Ref. 3) for subcooled liquid conditions, and Moody's analytical / experimental three pipe segment vessel blowdown with saturated steam conditions (Ref. 4).
The Joukowski analytical solution for instantaneous valve closure (Ref. 5) was also evaluated. REPIPE results provide excellent agreemert with all these problems, provided that recommended modeling guidelines are followed for both REPIPE and for RELAP5. Additionally, the EPRI/CE test 908 with a Crosby 6M6 L
safety valve and cold water loop seal has been analyzed with the RELAPS and REPIPE programs. Details of this analysis and comparison of REPIPE to test results and to other force post-processor methods are presented in the Validation Manual and also in the attached paper (Ref. 6, Attachment B), "Comparison of Analytical and Experimental Results for a
PWR Pressurizer Safety Valve l
Discharge,"
presented at the Third Multiphase Flow and liest Transfer r
Symposium-Workshop, Miami Beach, in April 1983. While the piping geometry with a t
cold loop sea! does not depict the Calvert Cliffs pressurizer safety and relief line piping configuration, this benchmark does confirm the RELAPS/REPlPE l
methodology and modeling approach. Benchmarks have also been performed for RELAPS/REPIPE nainst Moody's graphical solution for blowdown force resulting i
i r
l l
A*ITACliMENT A RESPONSE FOR REQUEST FOR ADDITIONAL INFORMATION TMI ACTION ITEM li.D.I PERFORMANCE TESTING OF RELIEF AND SAFETY VALVES from saturated steam source conditions, with an upstream restriction in the pipe (Ref. 4), again with excellent agreement.
states the safety valve flow area input to 8A, B G & E't October 23, 198{ supmittal was 1.4 x 10~
ft resulting in a ma f
rate of approximately lbm/hr. EPRI reported an area of 1.77x10"3s ft}ow RELAPS and a mass flow rate of 297,000 323,000 lbm/hr. Justify use of the smer flow area,
RESPONSE
Prior to performing the Calvert Cliffs safety and relief valve piping analysis, a RELAP5/ MODI model was developed to simulate the EPRl/CE test for the Dresser Model 31739A safety valve with steam inlet conditions. The objective of this model was to establish an appropriate RELAP5 safety valve model to match the test pressure and flow conditions. This model was then used for the Calvert Cliffs piping gonfiguratiog ang), inlet pressure. The valve bore area was given as 2.545 in (1.77x 10~
ft' but the actual valve discharge coefficient was not known. The safety valve was modeled in RELAP5 using the motor valve option, with the single-phase abrupt area change model to account for vena contracta losses at the throat. No external losses (due to the valve geometry) were applied. Use of the actual bore area with no external loss resulted in over-prediction of the test flow rate witu steady state inlet pressure of approximately 2600 psia. The effective valve throat area was reduced accordingly to match the test f'ow rate as closely as possible. The resulting effective area is in agreement with that determined from the compressible, critical flow equation, or that determined using Moody critical mass flux (Ref. 7).
The same valve model was used for the Calvert Cliffs safety and relief valve piping analysis. The resulting mass flow rate of approximately 297,000 lbm/hr was lower than the test flow rate, since the pressurizer pressure used for the valve inlet condition was lower than that of the test.
from 2.5x10'4 to 8 B.
The tige step used in the piping analysis was said to vary 5.0x 10-seconds. To prevent the shock wave generated by a valve discharge fronj passing through a volume in one time step the time step should be 1.265x10~
seconds. Justify the time steps usad in the RELAPS analysis.
RESPONSE
As stated in the initial request for additional information (RAI) response, the RELAPS time step used in th Cliff and relief valve piping 2.5x10~g Calvert5.0x10') safety analysis varies between and seconds. This information was based solely on RELAPS major edit summaries of the time step control convertence.
lx10-7 seconds. A small The specified minimun time step for all cases was percentage of the total time step advancements used less than 2.5x10-4 seconds.
Th? automatic time step control scheme of ?tELAPS (Ref. 8) used several criteria ATTACHMENT A RESPONSE FOR REQUEST FOR ADDITiONALINFORMATION TMI ACTION ITEM II.D.I PERFOP.M ANCE TESTING OF RELIEF AND S A FETY VALVES fo. assessment of the adequacy of the selected tlme step. For the saturated steam conditions appropriate for the Calvert Cliffs analysis, the two significant criteria are the acoustic or mass transport Courant limit (represented simply by At=
%x, where 'x' is the minimum control volume length and
'a' is the using the st sonic speed) and the density difference between that calculated time step (5x10'gte equations. If either is not satisfied with the maximum input
),
the time step is halved and the evaluation repeated. The process is repeated unti! '. O cot..ergence criteria are satisfied or the minimum input time step I
(1x10 seconds) is reached. If th criteria are still not satisfied, the time step is f;agged as unsuccessful and the run proceeds to the next transi(nt iteration step. All of the RELAPS runs for the safety and relief valve transient analysis indicated an insignificant number of repeated or unsuccessful advances.
The potential for supersonic shock valve transition was not expected to occur in the small piping upstream of the valves (since th. geometry is not conducive to supersonse flow), t ut was considered for the larger downstream piping. Allowing for the larger dit. meter and corresponding increased node length, the 2.S x 10-second time step is small enough to account for supercanic shock transition, if predicted to occur. This approach is consistent with Reference 9.
Supersonic velocities were predic.ed by RELAP5 nt enlarging tees in the downstream pipe during the transient, but none were observed in upstream junctions.
Additio. ally the input minimum and maximum time steps used are consisteat with those useo is RELAP5/REPIPE benchmarks (see response to Quea: ion #7).
9.
The structural analysis was performed using Bechtel's computer program, ME-101.
Provide more details on the verification of M E-101, including comparisons of calculated results and EPRI/CE data for our review.
RESPONSE
The ME-101 Program has been verified against the following standard piping
- programs, o
ME-632 o
EDS SUPERPIPE o
NUPIPE o
TRIPE o
ADINA o
MSC/ NASTRAN o
EASE 2 o
ANSYS o
PIPESD o
Pressure vessel and piping 1972 computer programs verification, The American Society of Mechanical Engineers
ATTACllhiENT A RESPONSE FOR Ri! QUEST FOR ADDITIONA L INFOR M ATION TMI ACTION ITEM II.D.1 PERFORMANCE TESTING OF RELIEF AND SAFETY YALVES The comparison of calculated results of force time history analysis is attached with this response for your information (see Attachment C). The M E-101 verification manual it considered proprietary to Bechtel. However, t' s manual can be made available for further review by the Nuclear Regulatory Commission at any Bechtel office upon request.
10.
Mo.e inforn.ation on the integration time step used in the structural model is needed. Also, provide information on the structural model lumped mass spacing
- nd the damp..
- g factors used in the analysis. Justify that all of the desir3d structural frequencies would be accounted for in the structural response with these modeling techniques.
RESPONSE
The forcing function generated by RELAP5 and REPIPE was carefully reviewed before selecting the time steps and cut-off frequency. The time steps of 0.001 second was chosen so that all the peaks in the force time !i.itory were accounted for as accurately as possible. Further the forcing function was reviewed and all significant contributury modes and their frequency contents were evaluated and it was determined t'iat at a cut-off frequency of 100HZ, all the significant harmonics will be accounted for. The dynamic analysis was carried out to include all harmonic responses up to 100 H Z.
The total response of the system was calculated by superposition of the esponse from all modes up to 100 11Z. The time history analysis w5 t performed using a conservative one percent of critical damping. The lumped masses are c 'refully located to adequately represent
^e dynamic properties or the pipini otem. A lumped mass is located at the beginnlag and end of every elt c,
valve, at the extended valve operator, intersection of every tea or branch connection and at least two mass points between two supports in the same direction. There is a total of approximately 450 ft of pipe per unit, from pressurizer nozzle to the quench tank, represented by 333 lumped mass points.
11.
Provide a discussion on w he t'.m additio-! moisture from the pressurizer spray was considered in determinir vc.c.. mlet conditions at d in the analysis done to select the transient producu.g :he maximum loads on tne safety valve discharge pipiag.
1
RESPONSE
The operation of pressurizer spray will not increase the discharge piping peak loads because the peak load occurs prior to the time when any wet steam due to ent:ained spray can reach the safety valve.
' n
A'ITACIIMENT A RESPONSE FOR REQUESTFOR ADDITIONALINFORMATION TMI ACTION ITEM II.D.1 PERFORMANCE TESTING OF RELIEF ANDSAFETY VALVES The msximum discharge piping loads occur upon valve opening. This means that the peak force in a given piping segment occurs when the initial pressure surge due to valve opening reaches that segment. The inlet piping for the safety valve will inith*1y contain saturated steam. In order for any postulated wet steam to reach the discharge piping, the initial quantity of saturated steam must pass through the safety valve. For Calvert Cliffs Units 1 and 2, this would take at least 0 015 seconds after the valve initially opens. By the time the postulated wet steam reaches the valva, the valve is fully open and the initial pressure surge has already c.,:urred. This is further substantiated by EPRI safety valve test data for si. m-to-water transition tests. In these tests the safety valve actuated on saturated steam followed by a transition of saturated water after the valve opened. The peak loads occurred when the valve initially opened prior to the transition to wa ter. Therefore, the operation of pressurizer spray will not result in discharge piping loads in excess of those valties previously presented in the Calvert Cliffs safety valve report.
In addition to the ASME Code report, the EPRI test program demonstrated the strt'ctural adequacy of the safety valve during valve actuation transients.
Also, the bending raoments predicted to act on the safety valve discharge flange in the Calvert Cliffs piping analysis arc less than those measured during the test program. The operability of the safety valves is therefor.: r.ot impared by the calculated piping loads.
12.
Pressurizer nozzle loads during safety valve and PORV discharge were not discussed. Compare the calculated and allowable loads for the pressurizer nozzles.
{
RESPONSE
1he fohawing table compares the CE allowable loads (Ref. II).with the calculated loads for all four nozzles. In all but one esse the calculated loads are considerably lowei than the CE allowable. The "Fy" load for nozzle number two for pressurizer eleven is slightly higher (2,6%) than the allowable. however, the resultant force is considerably less than the rertitant allowable and it was considered acceptaole. The calculated loads include the loads due to dead weight, thermal, seismic, seismic anchor movement and PORY and safety discharge loads.
r 9
_9
ATTAC11 MENT A RESPONSE FOR REQUEST FOR ADDITIONALINFORMATION TMI ACrlON ITEM II.D.1 PERFORM ANCE TESTING OF RELIEF AND SAFETY VALVFS Calculated / Allowable Load Calculated Loads Calculated Loads Allowable Pressurizer No.
11 Pressurizer No.
21 Loads Nozzle Nozzle Nozzle Nozzle No. I No.2 No. I No.2 FX (Ibs) 1010 1170 1490 2120 5472 FY (Ibs) 6220 6920 3120 4440 6743 FZ (lbs) 3390 3780 3310 2550 5472 Mx(ft-lbs) 3010 3430 1560 3010 9208 My(ft-lbs) 2120 1610 3340 2210 8859 Mz (ft-lbs) 2560 3470 1560 2200 9208 13.
In your response to Question 13 of our initial RAI, it was stated the Class I piping was analyzed using the given load combinstions and the USAS B31.7 1969 Code as indicated by a series of referenced Code sections and equations. This information was not specific enough to allow identification of the load combinations and allowable stresses for the Class -- I piping without considerable effort. Therefore, provide the specific load combinations and code allowables i
used in the Class I piping stress analysis.
RESPONSE
The followmg loading combinations and stress allowable were used for the Nuclear Class I piping.
1 j
ATTACIIMENT A RESPONSE FOR REQUEST FOR ADDITIONAL INFORMATION TMI ACTION ITEM II.D.I PERFORMANCE TESTING OF REI.IEF AND SAFETY VALVES Plant / System Operatirg Load Allowable Combinations Condition Combination Stress 1
Normal N
1.5 Sm 2
Upset N+SOTU+OBE 1.5 Sm 3
Emergency N+ SOT 1.5 Sm 4
Faulted N+DBE+ SOT 3 Sm E
5 Faulted N+DBE+ SOT 3 Sm U
N
= Sustained loads during normal plant operation Relief valve discharge transient SOT
=
U Sa ety valve discharge transient r
SOT
=
E Operating basis earthquake OBE
=
=
Allowable design stress intensity value at operating temperatures Sm
=
Also, as indicated in response to Question #13 of the initial RAI, all Nucleur Class 1 piping was analyzed using the above loading combinations and USAS B31.7 1969 Code as indicated below.
o Primary stress intensity limit for each point analyzed in accordance with Code requirements 1-705.1 (Equation 9) usinj the allowable indicated in above table, o
Primary plus secondaiy stress intensity range for each point analyzed in accordance with Code requirements 1-705.2 (Equation 10) and 1-705A (Equation 13 and 13).
o Cumulative damage for each point analyzed in accordance with Code requirements 1-705.3.4. i
NTTACllMENT A RESPONSE FOR REQUESTFOR ADDITIONALINFORMATION TMI ACTION ITEM II.D.I PERFORMANCE TESTING OF RELIEF AND SAFETY VALVES 14.
In response to Question 14 of our initial RAI, it was stated the maximum usage factor for the Class I piping was 0.0084 versus an allowable of 1.0. Since the usage factor only compares the number of actual cycles to the number of design cycles, this number does not provide any information on the calculated versus allowable stresses for the Class I piping. Provide a table for the Class I piping compariad the calculated and allowable stresses for the most highly loaded locations.
RESPONSE
The following table summatizes the five highly loaded points and the stress values are compared against the code allowable for Class I pipirig.
Data Eauation 9 Stresses Ecuation 10 Stresses Point Calculated Allowable Calculated Allowable Stress Stress Stress Stress (ksi)
(ksi)
(ksi)
(ksi) 4A 12.77 24.64 32.1M 49.275 12 14.70 24.64 40.667 49.275 41 9.53 24.64 36.798 49.275 43 11.80 24.64 32.634 49.275 37 9.53 24.64 39.665 49.275 15.
The allowable stresses for the Class I and Il piping supports were given as a fraction of the minimum yield stress. Provide the specific reference in the USAS B31.7 1969 Code that defines 'he al!owable support stresses in this way.
RESPONSE
The piping supports addressed in response to Question #13 of the initial R AI, consist of manufacturer's standard components attached to piping, such as struts and snubbers, as well as structural steel rolled sections used as supplementary steel to transfer loads to the building structure. The evaluation of standard components was based on the manufacturer's published load ratings. The structural rolled sections were evaluated using stress allowables based on the American Institute of Steel Construction (AISC) standard practice as prescribed by Section 1-720.2.4 of the USAS B31.7 1969 Code.
12 -
ATTACilMENT A
- RESPONSE FOR REQUEST FOR ADDITIONAL INFORM ATION TM1 ACTION ITEM II.D.1 PERFORM ANCE TESTING OF RF LIEF AND S AFETY VA LVES Section 1.5 of the Specification for the Design Fabrication and Erection of Structural Steel for Buildings, as found in the Eighth Edition of the AISC Steel Constructior, Manual, defines allowable stresses as a fraction of the minimum yield stress. This methodology is coma.dy used in the industry.
16.
Provide
- e. formation to show the support modifications required to reduco stresses to withu code allowables were completed. If the required modifications have not yet been made, provide a schedule outlining when this work will be complete.
RESPONSE
The required support modifications were completed.
1 17.
Bechtel's report on the discharge piping system identified the conditions analyzed for the PORY as max pressure, 2538 psia, and pressure ramp rate, 46.0 psi /sec. The safety valve conditions were 2534 psia and 64.4 psi /sec for the maximum pressure and pressure ramp rate, respectively. In BG&E's response to our initial RAI the conditions analyzed were identified as 2434 psia, maximum pressure, and 46.0 psi /sec, pressure ramp rate, for the PORVs. The safety valves were analyzed for 2538 psia and 64.4 psi /sec. Identify the actual conditions analyzed for Calvert Cliffs I and 2.
RESPONSE
The appropriate boundary conditions for the Calvert Cliffs safety and relief valve opening transient are identified in the table below. A small conservatism was incorporated in the RELAPS safety vahe model, ;6 that a maximum pressure of 2538 psia was applied for both PORV and Gafety valve opening transients. The response to Question 10 of the initial RAI suboitta) incorrectly reported the maximum PORV pressure used in the analysis.
Transient Valves hfax. Pru R_amo Rate Loss of Load PORV 2538 psia 46.0 psi /see Loss of AC Safety 2534 psia 64.4 psi /sec (2532 actually used) 18.
The CE inlet conditions report listed the FSAR transients and accidents for each plant which result in a peak pressure greater than the safety valve setpoint.
For some plants, this list included the feedwater line break (FWLB), but for other plants the FWLB was not included. Calvert Cliffs 1 and 2 was a plant that did not include the FWLB in its list of transients and acektents that challeng the safety valves. From the CE report it was not clear whether the FWLB was missing because Calvert Cliffs was licensed prior to the issuance of Reg.
' J
... ~
m
61TACIIMENT A RESPONSE FOR REQUESTFOR ADDITIONALINFORMATION TMI ACTION ITEM II.D.1 PERFORMANCE TESTING OF RELIEF AND SAFETY VALV FS Guide 1.70, Rev. 2 and, therefore, the FWLB was not initially analyzed as part of Calvert Cliffs' design basis. Discuss why the FWLB was not listed for Calvert Cliffs. If the FWLB was not listed for the second reason discussed above, it is the staff position that the Calvert Cliffs submittal is incomplete. I.em II.D.1 in NUREG-0737 specifically requires that PORVs and safety valves be qualified for fluid conditions resulting from transients and accidents referenced in Reg.
Guide 1.70, Rev. 2. Additionally, from the staff review of other plant-specific responses to item II.D.1, it is clear that the many plants the FWLB accident is the limiting case for providing high pressure liquid to the safety valves, a fluid for which they were not specifically designed originally. This is exactly the type of concern that NUREG-0737, ll.D.1, was established to address. In accordance with the requirements of the NUREG, we require that information be provided to demonstrate that tl'e PORVs and safety valves will function as required to assist in safe shutdown of the plant and will not experience any degradation that would inhibit safe plant shutdown if exposed to the FWLB.
RESPONSE
Feedwater line break was missing from the CE report because, at the time, FWLB was not considered a design basis event and wc understood the requirement of item II.D.1 to address design basis events in each plant's FSAR.
In conversations with NRC Staff subsequent to your question, it was explained that the intent of Itam II.D.]
is to address the transients and accidents referenced in Reg. Guide 1.70, Rev. 2, regardless of whethe-these transients and accidents are in our FSAR. T1.erefore, we should consider FWLB whether it is a design basis event or not. We accept this position and will respond no later than March 31, 1988, concurrent with the related open item: Is feedline break a design basis event at Calvert Cliffs? We will demonstrate the operability of the PORVs and safety valves at that time, regardless of the answer to that question.
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