L-87-448, Responds to NRC 870610 Request for Addl Info Re TMI Item Ii.D I of NUREG-0737, Performance Testing of Relief & Safety Valves. Remaining Questions Will Be Addressed by 880205
| ML17221A498 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 11/06/1987 |
| From: | Woody C FLORIDA POWER & LIGHT CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM L-87-448, NUDOCS 8711100388 | |
| Download: ML17221A498 (12) | |
Text
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ACCESSION NBR:
FAC IL:50-335 50-38'P AUTH. NAME MOODYI C. O.
RECIP. NAME REGULATOR NFORMATION DISTRIBUTION TEM (RIDS) 8711100388 DOC. DATE: 87/1 1/06 NOTARIZED:
NO St.
Lucie PlantI Unit 1I Florida Power 5 Light Co.
St.
Lucie PlantI Unit 2I Florida Power 8c Light Co.
AUTHOR AFFILIATION'loridaPower 5 Light Co.
RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
DOCKET 0 05000335 05000389
SUBJECT:
Responds to NRC 870610 request for addi info re TMI Item II D I of NUREQ0737I "Performance Testing of Relief Sc Safety Valves. " Remaining questions will be addressed bg 880205.
DISTRIBUTION CODE; AO46D COPIES RECEIVED:LTR J ENCL J SIIE:
TITLE:
OR Submittal:
TMI Action Plan Rgmt NUREQ-0737 5 NUREQ-0660 NOTES:
RECIPIENT ID CODE/NAME PD2-2 LA TOUR IGNYI E INTERNAL: AEOD/DOA ARM/DAF/LFMB NRR/DEST/ADS NRR/DREP/EPB N
S/ILRB 01 RES/DE/EIB EXTERNAL:
LPDR Nslc COPIES LTTR ENCL 1
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1 1
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REC IP IENT ID CODE/NAME PD2-2 PD AEOD/DSP/TP*B NRR/DEBT/ADE NRR/DEST/MEB NRR/DREP/RPB OQC/HDS2 RES DEPY QI NRC PDR CQP IES LTTR ENCL 5
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L7-448 NOVEMBER 0 6 1987 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen:
Re:
St. Lucie Unit Nos.
I and 2 Docket Nos. 50-335 and 50-389 TMI Action Item I I.D. I Re vest for Additional Information By letter dated June IO, l987 (E.
G. Tourigny to C. O.
Woody),
the NRC identified additional information the staff required to continue its review of TMI Item I I.D.I of NUREG - 0737,
'Performance Testing of Relief and Safety Valves".
By letter L-87-339 dated August l4, l987, Florida Power
& Light Company (FPL) provided a completion date of November 6, l987 for response to St. Lucie Unit I questions I., 2.A., 2.B., 4.A. and 5., and St. Lucie Unit 2 questions I.A., I.B., 2.A., 2.B.I., 2.B.2. and 2.B.4.
The purpose of this letter is to submit the response to these questions.
As stated in FPL letter L-87-339, the remaining St. Lucie Unit I and Unit 2 questions will be addressed by February 5, l 988.
If there should be any questions regarding this subject, please contact us.
Very truly yours,
~
~
C. O.
dy Group ice President Nuclear Energy Department COW/MSD/gc Attachment cc:
Dr. J. Nelson Grace, Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, St. Lucie Plant r 87iii00388 87i'i06 PDR ADOCK 05000335, p
PDR'pb MSD2/006/I an FPL Group company
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ADDITIONALQUESTIONS ON ST-LUCIE 1 SUBMITTAL Page 1
What is the torque setting (value and ft-lbs) used for the plant Limitorque block valve operators?
~Res onse The St.
Lucie Unit 1 motor-operated valve torque switch setting document specifies a nominal torque switch setting of 1.25 and a
maximum setting of 2.75 for the PORV block valves.
According to Limitorque, these torque switch settings correspond to a torque value of 47.5 ft-lbs and 98 ft-lbs, respectively.
2 ~
Insufficient detail was received on the key parameters used in the RELAP5/MOD1 thermal-hydraulic analyses.
Additional information is needed on the following items:
A.
B.
Node Size:
In Reference 1
the control volume
. size recommended is 0.5 to 1.0 ft. to adequately predict the fluid-hydraulic transient using RELAP5/MOD1.
What control volume sizes were used in the St.
Lucie 1 analysis?
If larger than recommended in Reference 1', verify that the model used predicts accurate or conservative loads.
Time Ste Size:
What calculational time step size was used in the analysis?
Reference 1
(page 2-6) states that the time step recommended was determined by dividing the shortest downstream control volume length by the estimated shock wave velocity based on an instantaneous valve opening.
The maximum shock wave velocity was assumed to be 2500 ft/sec (Reference 1,
page C-23). If a larger time step was used verify that its use produces accurate or conservative results.
~Res ense'A.
The pipe components immediately following the SRVs are component 5,
component 9
and.
component 17.
Component 5 consists of 18 volumes; component 9,
10 volumes and component 17, 15 volumes.
The following table lists the length of each volume in these 3
components:
Com onent, 5
Com onent 17 1
2 3
4 5
6 7
8 9
1.1 1;1 1.25 1.25 1.375
, 1.375 1 ~ 0.
1.0 1.0 1
2 3
4 5
6 7
8 9
1 '1.0 1.0 1.0
- 0. 65 1 ~ 0 1.0 1.0 1.0 1
2 3
4 5
6 7
8 9
1.0 1.0 1.0
'.0 1.0
, 1.0 0.72 0.8 0.8
Com onent 9
Page 2
Com onent 17 Vol. No.
Len th ft 10 11 12 13 14 15 16 17 18 1 ~ 0 1 ~ 0 0 ~ 91 0'5 0.58 1 ~ 0 1 ~ 0 1 ~ 0 1 ~ 0 10 1.0 10 11 12 13 14 15 0.9 1.0 1.0 1.0 1.0 1.0 The pipe components immediately following the PORVs are component 32 and 35.
The sizes of control volumes in these 2 components are listed as follows:
Com onent 32 Com onent 35 1
1.5 2
1.53 3,
3.67 4
4.25 5
5.0 6
5.0 7
-6.49 8
2.58 9
1.79 10 0.46 11 0.75 12 1.28 1
2 3
4 5
6 7
8 9
1.56 4.83 4.25 5.5 5.75 5.01 2.58 2.07 0.46 The volume sizes in the SRV lines are in general consistent with the EPRI guidelines.
The volume sizes in the PORV lines are larger than the recommended 1 ft.
length.
The PORU actuation transient is therefore less severe than the SRV actuation transient.
Consequently, the longer volume length in the PORV lines is tolerable and the overall RELAP5 model is still adequate.
Please note that the RELAP5 analysis for St. Lucie 1 was performed prior to the issuance of Ref.
1 by EPRI.
Xt is inevitable that a
slight deviation from the EPRI guidelines could be found in the model.
2B.
The time steps specified in the analysis range from 1.0 x 10-7 sec. to 2.0 x 10 sec.
The maximum time step recommended in Ref.
1 (page C-23) is 2.0 x 10 4 sec.
Therefore, the time steps used in the RELAP5 analysis are adequate.
I
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Page 3
provide the following information on the design analyses used to determine pipe stresses and support loads are within allowables:
What code or standard was the pipe stresses and support loads compared against to verify acceptability?
(ASME Section III, USAS B31.1 or ?)
If not clearly defined by the code
- used, what allowable stresses were used to compare with the predicted pipe stresses and support loads?
Show a comparison of the highest stressed and loaded areas with the allowable values.
~Res onse 4A.
5.
The portion of pipe from the pressurizer nozzles up to and including the safety and relief valves was analyzed in accordance with USAS B31.7 Class I, 1969 Code.
Although the remainder of the pipe up to the quench tank is classified as non-safety, it was included in the USAS B31.7 Class I piping model.
However, this piping was analyzed in accordance with ANSI B31.1, 1973 Code.
Safety related standard component supports were designed per USAS B31.7 Code and non-safety related standard component supports were designed in accordance with the ANSI B31.1 Code.
The codes used clearly specify the allowable stresses.
These allowable stresses were used to compare with the predicted pipe stresses and support loads.
The Combustion Engineering (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.
St.
Lucie 1 was a plant that did not include the FWLB in its list of transients and accidents that challenge the safety valves.
From the CE report it was not clear whether the FWLB was missing because the accident did not challenge the safety valves'r because St.
Lucie 1
was licensed prior to the issuance of Regulatory Guide 1.70, Rev.
2
- and, therefore, the FWLB was not analyzed as part of St.
Lucie design basis.
Discuss why the FWLB was not listed for St.
Lucie 1.
If the FWLB was not listed because of the second reason discussed above, it is the staff position the St.
Lucie 1
submittal is incomplete.
Item II.D.1 in NUREG-0737 specifically requires that PORUs and safety valves be qualified for fluid conditions resulting from transients and accidents referenced in Regulatory Guide 1.70, Rev 2.
The FWLB is specifically defined in Regulatory Guide 1.70, Rev.
2.
Additionally, from the staff review of other plant-specific response
.to Item II.D.1, it is clear that for 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, Item II.D.l was established to address.
In accordance with the requirements of the
- NUREG, we, require that information be provided to demonstrate that the 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.
Page 4
~Res onSe The feedwater line break (FWLB) event had been previously evaluated in Section 15.2.8 of the St.
Lucie Unit 1 FSAR.
Based on the evaluation assumptions presented in the
- FSAR, the FWLB event is a
cooldown event in the licensing basis for St.
Lucie Unit 1.
Standard Review Plan 15.2.8, Rev.
1 "Feedwater System Pipe Breaks Inside and Outside Containment (PWR)" states that depending on the plant initial conditions and assumptions, the FWLB could cause either a
reactor coolant system cooldown or heatup.
For St.
Lucie Unit 1, the FWLB event has been determined to be bounded by the limiting cooldown event, the main steam line break (MSLB) since the area for flow in the FWLB event is less than that assumed in the MSLB event.
As such, the pressurizer safety valves and PORV's would not be subjected to high pressure liquid discharge during this transient.
The Loss of Load event is still the most limiting plant heatup, or RCS pressurization event.
During this transient, the pressurizer safety valves and PORV's are limited to steam discharge.
Page 5
ADDITIONALQUESTIONS ON ST.
LUCIE UNIT 2 SUBMITTAL Insufficient detail was received on the key parameters used in the RELAP5/MOD1 thermal-hydraulic analyses.
Additional information is needed on the following items:
A.
Node Size:
In Reference 1
the control volume size recommended is 0.5 to 1.0 ft. to adequately predict the fluid-hydraulic transient using RELAP5/MOD1. 'hat control volume sizes were used in the St.
Lucie 2
analysis?
If larger than recommended in Reference 1 verify that the model used predicts accurate or conservative loads.
B.
Time Ste Size:
What calculational time step size was used in the analysis?
Reference 1 (page 2-6) states that the time step recommended was determined by dividing the shortest down-stream control volume length by the estimated shock wave velocity based on an instantaneous valve opening.
The maximum shock wave velocity was assumed to be 2500 ft./sec.
(Reference 1,
page C-23).
If a larger time step was used verify that its use produces accurate or conservative results.
~Res ense 1A.
The pipe.
components immediately downstream of the SRV's are component 5,
component 12 and component 19.
The sizes of all the control volumes in these three components are ranged from 0.58 ft.
to 1.0 ft.
The pipe components immediately downstream of the PORV's are component 33 and component 37.
The volume sizes in these two components are r'anged from 0.9 ft. to 1.0 ft.
/
Therefore, the volume sizes for piping downstream of the SRV's and PORV's are all consistent with the recommendations of Reference 1.
1B.
The time-step control specified in the RELAP5 analysis is described as follows:
Time Sec Minimum Time Ste Sec Max. Time Ste Sec 0 to 0.20
- 0. 20 to 0. 40 0.40 x 1.20 1.0 x 10 1.0 x 10 1.0 x 10 6
2.0 x 10 4
5.0 x 10 4
1-0 x 10
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Page 6
The RELAP5 program would try to use the maximum'ime step specified.
If the solution does not converge with the assigned time step, the program would automatically reduce the time step in half and repeat the calculation until the solution converges or until the minimum time step is used.
If the solution does not converge with the minimum time step, the program would stop the execution.
During the SRV/PORV actuation, the most severe transient generally occurs prior to 0.20 sec.,
therefore the time step control used in the RELAP5 analysis for St.
Lucie Unit 2 is adequate and consistent with the recommendation of Reference 1.
Provide the following information on the PIPESTRESS 2010 analyses used to determine pipe stresses and support loads are within allowables:
A.
B.
What code or standard was the pipe stresses and supports loads compared to show adequacy?
(ASME Section III, USAS B31.1 or ?).
If not clearly defined by the code used, what allowable stresses were used to compare with the predicted pipe stresses and support.
loads?
Show a comparison of the highest stressed and loaded areas with the allowable values.
The dynamic piping model used affects the accuracy of the predicted stresses and loads.
Provide the following information on the mode used:
(The figures provided were not adequate or legible).
1 ~
2.
4 ~
Maximum and minimum lumped mass spacing used.
Calculation time step used.
What damping factor was used in the analysis?
Typically 14 for upset and 2% for emergency conditions are the maximum allowed; if greater than these values were used, justification is requested.
~Res ense 2A.
The portion of pipe from the pressurizer nozzle up to and including the safety and relief valves was analyzed in accordance with Article NB-3600 of ASME Boiler and Pressure Vessel Code Section III, 1971
- Edition, including Summer 1973 Addenda.
Although the remainder of the piping up to the quench tank is classified as non-safety, it was analyzed using the Class 2
requirements of Article NC-3600 of ASME Section III, 1971 Edition including Summer 1973 Addenda.
This piping was upgraded to Safety Class 2 since it was included in the ASME Section III, Class 1
piping model.
Safety related standard component supports were designed per the requirements of ASME Section III, 1971 Edition, including Summer 1973 Addendum and non-safety related standard component supports were designed in accordance with the ANSI B31.1 Code.
The codes used clearly specify the allowable stresses.
These allowable stresses were used to compare with the predicted pipe stresses and support loads.'
4 "r
Page 7
2B1.
The maximum and minimum lumped mass spacing used in the analysis is as follows:
PIPE SIZE 3" Sch.
160 4" Sch.
160 6" Sch.
40 8" Sch.
40 10" Sch.
40 MAXIMUM 2.333 ft.
2.229 ft.
4.537 ft.
5.292 ft.
4.685 MASS PO NT SPACING MINIMUM 0.333 ft.
0.154 ft.
0 322 ft.
0.145 ft.
0.167 ft.
2B2. The calculational time step used in the dynamic analysis of piping utilizing the PIPESTRESS 2010 program are the same as those used for the other RELAP5/MOD1 thermal-hydraulic analyses as given in
Response
1B.
2B4.
The damping factor used for the upset condition was 14 and for the emergency condition was 24.
REFERENCES 1.
"Application of RELAP5/MOD1 for Calculation of Safety and Relief Valve Discharge Piping Hydrodynamic Loads"i EPRI 2479/
December 1982.
VL/SUBMIT.RPT0001