ML20128C013
| ML20128C013 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 11/30/1992 |
| From: | Joshua Wilson TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9212040198 | |
| Download: ML20128C013 (22) | |
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4 w.,# u nw m. 9 November 30, 1992 U.S. Nuclear Regulatory Commission ATTNt Document Control Desk Washington, D.C. 20555 Gentlement In the Matter of
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Docket Nos. 50-327 Tennessee Valley Authority
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50-328 SEQUOYAll NUCLEAR PIANT (SQN) - MAIN STEAM VALVE VAULT DESIGN CONCERN AS A RESULT OF NONCONSERVATIVE MASS AND ENERGY RELEASE DATA FROM MAIN STEAM-LINE BREAK ANALYSIS - JUSTIFICATION FOR CONTINUED OPERATION (JCO)
Reference:
Licensee Event Report dated August 17, 1992, " Tennessee Valley Authority - Sequoyah Nuclear Plant Units 1 and 2 -
Docket Nos. 50-327 and 328 - Facility Operating Licenses DPR-77 and DPR Licensee Event Report (LER) 50-327/92013" As discussed in the above reference, on July 15, 1992, it was-determined that the mass and energy releases used to calculate the pressures, which would be experienced during a main steam line break in the main steam valve vaults, were not conservative. A JC0 was su'usequently: developed and provided to the NRC resident inspectors and NRC Nuclear Reactor Regulation (NRR) staff. A subsequent telephone conference with both Region II and NRR staff on July 17-identified additional NRC questions.
A response to these questions was provided to NRC on July 24 TVA was verbally informed that no further concerns were' identified as a result of
~
NRC review of the provided information.
A verbal request from the NRC staff was received on November 4, 1992, to formally docket the previous information that had been provided for staff review.
The enclosure provides the most recent revision to the JC0 for the subject deficiency.. Please note that a revision to the above reference is planned in April 1993 to provide a status of the Engineering evaluation efforts.
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O.S. Nuclear Regulatory Conenission Page 2 November 30, 1992 Please direct questions concerning this issue to J. D. Srnith at (615) 843-6672.
Sincerely, 1
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/J L. Wilson Enclosure cc (Enclosure):
Mr. D. E. LaBarge, Project Manager U.S. Nuclear Regulatory Coniniission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 NRC Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy Tennessee 37379-3624 l
Mr.
B.'A. Wilson, Project Chief U.S. Nuclear Regulatory Consnission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Gaorgia 30323-0199 i
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ENCLOSURE
,1 JUSTIFICATION FOR CONTINUED OPERATION FOR MAIN STEAM VALVE VAULT DESIGN CONCERN 4
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JC0 92-011, R2 j of 21 SITE JUSTIf! CATION FOR CONTINUED SSP-12.14 STANDARD OPERATION / WAIVERS Of Rev. O PRACTICE COMPLIANCE Page 19 of 54 APPENDIX C ENGINEERING EYALUATION/JUTTIFICATION FOR CONTINUED OPERATION (JCO)
(JCO) # sg Jco 92-011, Revision 2 Corrective Action Document #
11-5-92-059 Engineering Evaluation 0 JC0 (1 JCD/ Engineering Evaluation including actions to resolve (Note:
Actions /special conditions shall be tracked by the associated Corrective Action Document (s)
See attached docunent for justification for continued operation.
This JCO is effective through the Cycle 6 refueling outage for both units.
Prepared by:
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D. M. Latever' Reviewed by:
O V. A. Bianco Cross Discipline Review ANY 9
2.
NE - Civil hAUbvn~u pli1/U-NE - Materials #
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Site Engineering / Tech Support Duty'Plar;t Mana r/ Operations Management Attach copy to corrective action document Site VP concurrence to exceed time requirements O
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SiteVicePresidentMr/
of Section 3.3.1 j/..........................9t Required actions are complete and JC0 :an 3e closeM forward copy to DCRM for removal frcxn controlleo manuals.
Responsible Org Nuclear Eng/ Tech Support Plant Manager /Date 08655/voo
JC0 92-011. R2 L of 21 N
EVALUATION OF TNE POTENTIAL TOR A FULL CIRCUMTERINTIAL PIPE BREAK IN THE MAIN STEAM PIPING IN THE VALVE VAULTS AT SEQUOYAH NUCLEAR PLANT DESCRIPTLQN Or ADVERSE CONDITIorf:
Calculation Nos. SQN-APS2-023 (West Valve Vault) and SQN-APS2-042/TI-295 (East Valve Vault) establish the structural design pressurns for the Sequoyah main steam valve vaults under the limiting design basis pressure transient (dou:ie ended severance of a main steam line inside the vaults).
These calculations use mass and energy release data submitted to TVA by Westinghouse Letter TVA-2230 dated November 8,1971. The information provided in Letter TVA-2230 was based upon the latgest steam generator depressurization rate consistent with dry steam discharge.
Subsequent review of a similar mass and energy release calculation for the Watts Bar Nuclear Plant has prompted Westinghouse to conclude that the mass and energy releases in Letter TVA-2230 may not represent the limiting break flov for vault pressure determination.
Westinghouse has indicated that the assumption of a dry steam release in a vented compartment such as the valve vaults is not conservative due to moisture entrainment within the discharge.
Westinghouse recommends in Letter TVA-92-133 that the methodology outlined in ANSI /ANS Standard 58.2 1980, Appendix E be used to generate bounding mass and energy release data for vault pressurization calculations.
The use of ANSI /ANS 58.2 methodology significantly increases the mass and energy releases presently used in the valve vau e pressure calculation. The increase in the mass and energy release has the potential to produce pressures which exceed present structural design margins and may challenge the structural adequacy of the reinforced concrete valls and slabs for the valve vaults.
7 DESCRIPTION OF THE VALVE VAULTS:
The valve vaults are located adjacent to the Unit 1 and Unit 2 containment building's.
Each unit has two valve vaults that are designated by their azimuthal direction from assumed North, i.e., East and West Valve Vaults (Tigure 1).
The East Valve Vault. of Unit 1 and the East Valve Vault of Unit 2 are identical. Similarly, the West Valve Vault of both units are identical.
The East Valve Vault is primarily a stand alone structure attached to the outside of the containment building;,whereas, the vest valve vault is sandwiched between the auxiliary building and the containment building with only one side open to the outside atmosphere. The valls of the valve vaults are constructed of reinforced concrete.
Steam lines and feedvater lines for two stekm generators are routed through each valve vault.
Several smaller high energy lines are also located in the valve vaults. A rupture in one of these high energy lines could result in either rapid pressurization of the valve vault and/or adverse environmental Page 1
JC0 92-011. R2 3 of 21 1
EVALUATION OF THE POTENTIAL FOR A WLL CIRCUMTERENTIAL PIPE BREAK IN THE MAIN STEAM PIPING IN THE VALVE VAULTS AT SEQUOYAH NUCLEAR PLANT (Continued) conditior.-.epending upon the break size, type, and location.
The valve vaults contain class 1E safety-related instrumentat..on and equipment necessary to mitigate the rupture of a main steam line or feedvater line in the valve vaults.
Blowout panels and roofs in the east and vest valve vaults are designed to relieve the pressurization transient resulting from postulated high energy line ruptures in the vault structure.
The primary function of the blovcut panela and roof was the integrity of the valve vault structure.
The flow areas created by the pressure relief panels and roof introduced atmospheric air into the lower elevations of the valve vault and exhausted the hotter air out the roof openings.
The duration and severity of the adverse MSLB temperature event iJ substantially reduced because of the availability of the pressure relief flow areas.
GliSEQUENCES OF THE ADVERSE CONDITION:
Postulated events potentially affected by the adverse condition are ruptures of the main uteam piping. The MSLB is the design basis event for the pressurization and temperature transients.
The MFLB is the design basis event for flooding in the valve vaults and is not impacted by this condition.
Therefore, postulated MSLBs are the only design basis event which must be examined.
For MSLBs in the valve vaults, the use of ANSI /ANS 58.2 methodology significantly increases the mass and energy releases presently used in the valve vault pressure calculation.
To quantify the effect of these increases on valve vault pressurization, an informal calculation was performed.
Based upon TVA Calculation CEB-PR/N2-091 R6, a main steam line break vas postulated to occur in the valve vault vall flued head anchor (refer to Figure 2).
(Breaks in the MS piping are required to be postulated at the terminal ends and intermediate points along the piping that exceed a stress limit of 0.8 (Sa + Sh). For this section of piping the stress limit is n-t exceeded and therefore the break only need be considered at the terminal end inside the flued head.) This anchor, along with the pipe restraint system (refer to figures 4 and 6), limited the break size to 3.36 ft2 (i.e., the annular area between the outside of the steamline and the inside of the anchor sleeves -
2 this area is significantly less than the 5.6 ft of a full guillotine pipe break).
The calculation found that the peak pressure in the valve vaults is less than the current design basis values.
However, the static pressure after the initial Page 2
JC0 92-011 R2 4 of 21 I
EVALUATION OF THE POTENTIAL FOR l
A FULL CIRCUMTERINTIAL PIPE BRIAK IN THE MAIN STEAM PIPING IN THE VALVE VAULTS AT SEQUOYA11 NUCLEAR PLANT (Continued) pac: cure peak is greater than the current design basis calculations.
The increased static loads could challenge the structural adequacy of the reinforced concrete valls and slabs of the valve vaults.
(See Attachment 1 detailed summary of the calculation and the results.)
s t axiure of the structural slab or valls could potentially damage equipment or i
piping in the valve vault associated with either the FW or ATV systems.
This could result in reduced flow or termination of flow to the steam generators.
Additionally, failure of the structural slab or valla could potentially damage the intact main steamline in the valve vault resulting in blevdown of a second steam generator.
These conditions could be outside the design basis of the facility.
TECHNICAL BASIS FOR CONTINUED OPERATION:
s Based upon NRC Branch Technical Position ASB 3-1 (Section B.1.a(1)), certain pipe breaks may be excluded from design basis consideration given that certain criteria are met.
For the purpose of this evaluation, the criteria for pipe break exclusion are contained in NRC Branch Technical Position MIB 3-1 (Section B.1.b).
While the criteria are stated to be applicable only to containment penetration areas, they are equally applicable to the SQN main steam piping from the steam generator, through the containment penetration and through the valve vaults.
TVA design requirements for all three areas are the same (i.e., Class B).
A review of the SQN design relative to the MEB 3-1 criteria was performed.
The following is a summary of that review:
B.1.b. (1).a through e - These criteria are for class 1 piping.
The piping in the valve vault is Class 2.
B.1.b(1).d - Based-upon a review of TVA Calculation.CEB-PR/N2-091 R6 (B41 900717 005), there are no' stress related breaks for the main steam piping. The main steam piping in the valve vaults meets this criteria.
B.1.b.(1).e - Since the piping is isolated from other piping by the flued head' anchor and the steam generator, no-loadings resulting from postulated piping failure vill result in stresses in excess of the 1.8Sh limit.
Page 3
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JC0 92-011. R2
~*
.5 of'21 EVALUATION OF THE POTENTIAL FOR A FULL CIRCUMFERENTIAL PIPE BREAK IN THE MAIN STEAN PIPING IN THE VALVE VAULTS AT SEQUOYAH FUCLEAR PLAJU (Continued)
B.1.b.(2) - Welded attachments (lugs) were utilized for some of the supports ou the main steam lines (refer to' table 2 f or a sammary of the velded attachments).
These lug locations were revieved to ensure the stress criteria of B.l.b(1) was met and vere found Yo meet the criteria.
The largest equation 9u plus equation 10 stress with the local lug iuduced stresses vas found to be less than the allowed 0.8 (1.2Sh + Sa) of 34,800 psi.
B.1.b.(3) - To the extent possible the number of circumferential and longitudinal piping velds and branch connections have been minimized (refer to table 1 for a summary of the circumferizatial velds).
There are no guard pipes in the area of the valve vaults. The piping an the containment penetration guard pipe in the annulus area meets the requirements of this element.
B.1.b.(4) - The geometry of the piping is such that the length of the portions of pipe have been minimized to the minimum practical length ~
(refer to Table 3 for a summary of pipe lengths).
B.1.b.(5) - The connections to the containment and to the anchor in the valve vault vall utilize flued integrally forged pipe attact -
and therefore meet the requirements of this element.
B,1.b.(6) - This item is not app'icable to this configuration. The design does not include. guard pipe or guard pipe assemblies for pressure retaining purposes.
The guard l
f pe over the main steam piping in the cautainte.* annulus area is an open ended element with no structural integrity function.
It does not preclude inspection of the enclosed main steam piping.
B.I b.(7) - A 100 percent volumetric inservice examination of all pipe velds has not been conducted.
Howev(r, a 100 percent inspaction of all velds was performed during original insttiletion (as van required of all TVA Class B tJping).
Additionally, this piping is included in the TVA in-service Lnapection program.(ASME Section XI, 1974 Edicion, Summer 197; Addenda).
Selected welds have been in-service inspected by TVA (refer to Tables 1 and 2 for a sumaary of the velds inspected). No prob 1 ma have been found with the velds inspected to date.
Since the in-service inspection program does not require 100 percent inspection each operacional cycle, the EPRI developed program CHECMATE was used to estimate the erosion / corrosion rate of the main steam piping and Page 4
'JC0 92-011, R2 k of 21 EVALUATION OF THE POTENTIAL FOR A FULL CIRCUMFERENTIAL PIPE BRIAK IN THE MAIN STEAM PIPING IN THE VALVE VAULTS AT SEQUOYAH NUCLEAR PLANT (Continued) components from the steam generator no::la to the flued head where the main steam piping exits the valve vault.
A steam quality value of 99.75% was utilized in the program.
(Jhis is consistent with the expected minimum steam quality described in the Sequoyah TSAR Section 5.5.2.2.)
The calculated erosion rate at the vorst valve it.:o is 2.9 mils per year for life of plant and 1.? mils per year c'trra.. rate.
For the worst elbov location, the cros on rate is 2.0 mils per year for life of plant and 0.6 mils per year cu" ent rate.
(The difference in the life of plant and current rates is dte to the use of Morpho 11ne water during the past year to improve secondary site chemistry.)
treatment These erosion / corrosion rates are considered to be very tav.
The EPRI CHICMATE program has been used during the last two outages to predict the components in certain turbine building systems (Extraction Steam, Feedvater, Heater Drains and Vents, and Condensate) that are susceptible to vall thinning by crosion corrosion.
The locations predicted by CEECMATE have been examined by ultrasonica during the last two outages.
The EPRI recommended correlation between predicted and measured vall thickness should be 50%.
Our U2C5 data (129 components) shows that we are generally within this 50% correlation range between predicted and measured vall loss.
Since these velds are covered by an in-service inspection program, the results of the inspections performed to date indicate that the velds are acceptable and the expected corrosion / erosion rate is low, the SQN piping meets the intent of this requirement.
Based upon the above review, TVA meets the intent of the NRC Branch Technical Positions for the exclusion of a double-ended pipe severance in the main steam valve vault.
Under the NRC requirements, pipes which meet the exclusir criteria must be analyzed for a 1.0 ft2 break.
2 A 1.0 ft break limits the fs out of the break to values significantly below the double ended flow rates.
The smaller break size reduces fluid velocities and level swell in the steam generators to the point that appreciable entrainment of liquid in the ateam would not be expected.
For a limited break outside containment, flow through the break would be essentially equally fed by all four steam generators. Thus, the effective flow increase due to the Page 5
JC0_92-011, R2
'7 of 21 EVALUATION OF THE POTENTIAL FOR A FULL CIRCUMFERENTIAL PIPE Bl. AK IN THE MAIN STEAM PIPING IN THE VALVE VAULTS AT SEQUOYAd NUCLEAR PLANT (Continued) break in any one generator vould be limited to the flow out a 0.25 f t2 break.
Based on Moody critical flow for no line loss and a quality of 0, the flow rate through a 1.0 ft* hole at 1000 psia is 7970 lb/sec.
The enthalpy of this fluid would be lov.
The steam flow used to design the valve vault valls was greater than 10,000 lb/sec with a high enthalpy.
It can therefore be concluded that a one square foot break would not result in higher valve vault pressures than had been used in the original design of the vaults.
Additionally, a 1.0 ft2 break was recently evaluated by Westinghouse to support modification of the east and vest valve vaults under DCNs M06646A, M06647A, M06648A, M06666A and M06667A (Unit 1) and M06649A,' M06650A, M06651A, M06710A and M06711A,(Unit 2).
(The 1.0 ft2 break size was evaluated along with 0.8 and 0.9 ft' breaks to establish the break which maximizes the temperature inside the valve vault - refer to SQ-RPT 25.38, B88 910918 010).
The valve vault roof modification analysis established that the blow-off roofs opened instantaneous to the break and the valve vault pressure remained at atmospheric pressure throughout the transient.
While calculation SQ-RPT 25.38 does not specifically address moisture entrainment in the mass and energy releases used in the analysis, the mass flow rates associated with the 1.0 ft2 break are lov.
Experience in analyzing steam line breaks in the past has established that moisture entrainment is not a significant consideration for mass flow rates as low as those used in the 2
1.0 ft break analysis. Accordingly, it can be concluded that mass and energy releases used in the 1.0 ft2 break analysis are not significantly affected by moisture entrainment and the conclusion that the break does not increase the valve vault pressure above atmospheric (after the roofs have lifted) remains valid.
As such, the structural integrity of the valve vaults is not challenged by the 1.0 ft2 break.
CONCLUSION:
The complete rupture of a main steam line is an FSAR condition IV design basis event.
It is not expected to occur'during the 40 year life of the plant.
It is a drastic scenario which represents a limiting design basis occurrence.
A review of the main steam piping relative to the NRC criteria in Branch Technical position MEB 3-1 has concluded that the piping complies with the general requirements for exclusion of a double-ended pipe severance from the 2
design basis.
Under the NRC' criteria, only a 1.0 ft break need be 2
considered.
Past analyses have indicated that a 1.0 ft break in the valve vaults vill not challenge the structural integrity of the va've vault walls or floor slabs.
As such, continued operation is justified through the next refueling outage of each unit.
Page 6 PLO25N04--3151
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TABLE 1 JC0 92-011, R2-Jhof21
- CIRCidG'ERDfTIAL *dELDS UNIT 1 UNIT 2 Loco Valve Vault Total
- valve Vault Tetal.
1 10 17 8
17 2
8 16 7
18 3
9 20 9
21 ~
4 8
LO LQ LQ O
37 73 34 76-
- Total - includes piping in valve vaults and conta.inment INSERVICE INSPECTION PROGRAM To Date 40 yr 1st 10 yr-To in Valve-Samole
,Jamole Date To Go Vault U1 Circumferential welds 13 4
2.
2 2
U2 Circumferential welds 13
-4
.2-2-
1"
= Welds located on safety valve header piping 4
L PLO25N04--3168
'i JC0 92-011, R2 15 of 21 TABLE 2 A'I'IACIIMENTS
- IRITT.1 UNIT 2 Loco Valve Vault Tetale.
Valve vault IetA *=
1 5
10 5
15 2
4 12 4
15 3
4 14 6
15 4
1 11 2
li 21 52 24 61
.:ncludes thermal wells:
Unit i valve vault 15 Unit 2 valve vault 17 Unit 1 total **
33 Unit 2 total **
34
= Total - includes piping in valve vault and containment INSERVICE INSPECTION PROGRAM To Date 40 yr 1st 10 yr To in valve Samole s arnele Date To ce vault U1 Integrally welded supports 10 10 13 "
0 2
U2 Integrally welded supports 10 10 7
3 2
- TVA has exceeded the required sample size.
PLO25NO4--3168 l
T.
JC0 92-011, R2
/G of 21 TABLE 3 PIPE LENGTH, Trs; APPROIDA*E UNIT 1 wo UNIT 2 Loon valve vault =
Tot a l" 1
20'-2*
91'-6*
2 14'-5" 85'-9" 3
28'-1*
100*-8" 4
28'-11" 101'-8" does not incude riser and header to safety valves.
valve fault
- Total - includes piping in valve vault and containment but does not include riser and header in vault valves.
PLO25tiO4--3168
JC0 92-011, R2 l'?of 21 1
An mial'ysis w as performed to determine the pressurization of the valve vault followi ng a MSLB.
The only double-ended br'ak location (s) in the e
MSVVs are ins ide the anchor sleeves.
The size of the break is limited at this locat ion to the annular area between the outside of the steamli anc the inside of the anchor sleeve.
This annular area is 3.36 ft The mass and energy release (M&E) from the MSLB was conservatively.
calculated us ing a methodology consistent with that given in ANSI /ANS 58.2-1980 wh< le accounting for a gradual change in break ". 'w quality.
In ANSI /ANS 68.2 1980,. theliSLB break flow consists of steam at 100%
quality unti' one-half the initial steam inventory in the steam generators in depleted.
The ANSI standard then assumes a-homogeneous mixture of sneam and saturated water exits the break, the homogeneous mixture cons-sts of the initial steam generator secondary side inventory less one-hal ' the initial steam inventory.
Instead of the step change in break flow quality, used in the ANSI standard, a more gradual change in break flow quality is supported gy WCAP-8822 and wa used in determining
- he M&E for the 3.36 ft MSLB.
The calculated M&E is given below:
Steam FL ow Liquid Flow (hs - 1181 Btu /lbm)
(h
- 557.S Btu /lbm)
Time!seci now(1bm/seci TimeJ sec)
Flow (lbm/sec) 0.0 7731.
0.0 0,
7731.(1) 0.4 0.4 0
1933.
1.5 19950(2) 1.5 2.5 1005.
2.5 23150 4.0
- 727, 4.0 24210 6.0 704.
6.0 24210 The M&E, giv en above, was used as input into multinode subcompartment models of tha MSVYs.
The subcompartment model of the West MSVV consists of 7 nodes a nd is shown in Figure 1.
The subcompartment model of the East tiSVV co isists of 9 nodes and is shown in Figure 2.
The pressurizati )n of the MSVVs was calculated using the MONSTER computer program.
Thb pressure profile with the highest peak pressure is given in Figure 3 fortheWestMSVVandinFigure4fortheEastMSVV.
The peak pressure in the West MSVV is calculated to be 21.68 psia which is less than cu rent value of 23.66 psia.
The peak pressure in the East MSVV is calc alated to be 18.22 psia which is less than the current value of 19.1 psia 1
0
I
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JWEST YALVE: VAULT N00Alf2AT10N JCO 92-011,-a2
/S of e 21_
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.EL. 706 SCHEMATIC OF WEST VALVE VAULT.
LEGEND
@FLOWJUNCTIONX
=.,.
.:?): c c M h 1 8 T erE M T FIGURE.I 4
,h
'N
SEQUOYAH EAST VALVE VAULT NODALI 2 ATION Jco 92-011 n2 jQof21 15 110 (0UTSIDE ATN0 SPHERE)-
o ELEVATIONS A N' "."- N/e$ sIij-) #:.3f-R I-i [ 4' '%[
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