ML20197G483
| ML20197G483 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 06/01/1984 |
| From: | John Miller Office of Nuclear Reactor Regulation |
| To: | Withers B PORTLAND GENERAL ELECTRIC CO. |
| References | |
| TAC-42594, TAC-51953, NUDOCS 8406150251 | |
| Download: ML20197G483 (9) | |
Text
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Dc5 8]S-oh Docket No. 50-344 JUN 1IM DISTRIBUTION:
t Xocket File PMKreutzer NRC PDR CTrammell L PDR Gray File Mr. Bart D. Withers ORB #3 Rdg EReeves Vice President Nuclear DEisenhut LHulman Portland General Electric Company OELD RWright 121 S.W. Salmon Street ACRS-10 GBagchi Portland, Oregon 97204 Edordan MFields JNGrace
Dear Mr. Withers:
This is a follow-on to our letter to you of January 26, 1984 on the subject of containment purging and venting.
We have completed our review of your submittais of January 25, 1993 and January 24, 1984 regarding the new containment hydrogen vent system isolation valves. Our review indicates that these valves are qualified to close against the buildup of containment pressure in the event of a loss-of-coolant accident and that they are therefore acceptable for service. Our Safety Evaluation is enclosed (Enclosure 1).
We have also enclosed a generic evaluation of the radiological consequences of accidents while purging or venting at power (Enclosure 2). The generic evaluation applies to Trojan since our review indicates that isolation signal circuitry and the new hydrogen vent valves are reliable.
This completes multi-plant action B-24, " Containment Purging and Venting" for the Trojan facility, including technical specifications. Your cooperation in resolving this issue is appreciated.
Sincerely, Original signed by DEsMa k{)(James R. Miller, Chief Operating Reactors Branch #3 Division of Licensing
Enclosures:
1.
Generic Evaluation cc w/ enclosures:
See next page ah
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Portland General Electric Company Gary Johnston, Resident Inspecto,r cc:
U. S. Nuclear Regulatory Commission Trojan Nuclear Plant P. O. Box 0 Rainier, Oregon 97048 Robert M. Hunt, Chairman Board of County Commissioners Columbia County St. Helens, Oregon 97501 Donald W. Godard, Supervisor Siting and Regulation Oregon Ocpartment of Energy Labor and Industries Building Room 111 Salem, Oregon 97310 Regional Administrator Nuclear Regulatory Commission, Region V Office of Executive Director for Operations 1450 Maria Lane, Suite 210
' Walnut Creek, California 94596 f
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t i-TROJAN NUCLEAR PLANT DOCKET NUMBER 50-344 c
DEMONSTRATION OF CONTAINMENT PURGE AND VENT VALVE OPERABILITY 1.0 Requirement f
i Demonstration of operability of the contairment purge and vent valves, par-ticularly, the ability of these valves to close during a design basis accident is necessary to assure containment isolation.
This demonstration of operabil-ity is required by BTP CSB 6-4 and SRP 3.10 for containnent purge and vent valves which are not Sealed closed during operational conditions 1, 2, 3, and 4.
2.0 Description of Purge and Vent Valves Valve No.
Size (Inch)
I HB E -B F-M0-10005 8
i HBE-BF-M0-10006 8
HBE-BF-MO-10007 8
[
HBE-BF-MO-10008 8
[
HBE-BF-M0-10009 8
HBE-BF-M0-10010 8
'p HBE-BF-M0-10011 8
HBE-BF-MO-10012 8
The eight (8-inch) containment mini purge and hydrogen vent system butterfly valves at the Trojan Nuclear Plant are supplied by the Clow Corporation and replace 8 valves of the'same size that were originally installed. The four j
54-inch valves are maintained closed by tne licensee-during operational condi-tions.1, 2, 3, and 4 Actuators used with the Clow valves are Limitorque actuator nodel no.
SMB-00-15-M2BC of the electric motor type.
}
3.0 Demonstration of Goerability l
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Tne Portland General Electric Compig (PGE) in their submittals dated January I
25, 1983 and Janua ry 24, 1984 provided the operability denionttration i n forma-tion for the purge and vent valves in their Trojan Nuclear Plant.
i A Clow Corporation report attacntd to the January 24, 1984 PtE submitt41 con-f tains tne specifics for demonstration of valve operability.
i A.
Valve loads are predicted using model test data for the prediction of dynamic torque coef ficients.
The model tests take into account the ef fects of upstream and downstream piping elenents (elbows, tees, etc.1 including various separation di stances and orientations. Cons erva tive assumptions are used in the determination of dynamic torque with no credit taken for pressure ramp in containnent and no credit taken for backpressure due to downstream piping.
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, B.
The SMB-00-15-H2BC actuator provided with all eight Clow valves, has a rated torque of 26,400 in-lbs providing a large torque margin when com-pared to the worst case dynamic torque of 6,375 in-lbs and the seating torque of 10,000 in-lbs.
C.
The valve frequency and stress analysis are acconplished by comparing PGE's design requirements (PEI-TR-83-33) to an already performed worst case generic qualification for the Clow 8-inch Wafer Stop Valve (PEI-TR-83-24). Table 1 presents the comparison of design requirements for Trojan vs the generically qualified data.
This table shows that the loading and dimensional parameters for the subject valves are adequately enveloped by the generic nuclear qualification.
D.
Seismic Quali fication is demonstrated by stress and frequency analyses prepared by Patel Engineers of Hurtsville, Alabama in Report No.
PE I-TR-83-24 Rev. A and PE I-TR-83-33.
4.0 Evaluation 4.1 The January 24, 1984 suomi t tal f rua PGE wnicn includes the Clow Corp.
report entitled " Purge and Vent Valve OperaDility Qualification Analysis" contains information that demonstrates adequate torque and stress margins.
The Clow analysis assumes worst case postulated accident conditions using peak containment pressure taken from the LOCA containment response curves and tases no credit for ramp pressure rise, which the staf f finds acceptably conservative.
Dynamic torque coef ficients are derived from. a model test program performed for Clow by Dr. A. L. Addy of the University of Illinois using the Uni-versity's facilities. From the data base developed by the model tests, a coaputer program (Clow valve analysis -program - CVAP) is used to predict valve operating characteristics.
Table 2 taken from Clow's = analysis tabulates the worst case dynamic torque for the two 8-inch valves installed with in-plane elbows di rectly upstream.
The peak dynamic torque of 6,375 in-lbs, occurring at a 60 valve angle is used for actuator sizing and structural analysis.
4.2 PGE presents in Tables 3 and 4 (Jan. 24,1984 submi ttal ) the SMB-00 H2BC actuator torque and structural margins. The torque switch trip setting of 6.360 in-lbs is adjusted lower than the seating and dynamic torques to take into account the torque magnification produced by tne inertial erfects caused by the fast closure times of approximately 3 seconds.
Tae staf f concurs with this approach to preclude the possibility of overstressing as a result of inertial ef fects, particularly dur ing seating.
4.3 The valve frequency and stress analysis are accoiplished by cunparing PGE's design requirements to the worst case generic qualification analysis for l
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. Table 1 Comparison of Trojan Nuclear Specific Requirements to Generic Nuclear Qualification Data '(Reference 8.0, A1)
Loadings Generic Trojan Pressure Shell (psig) 285 285 Seat (psid) 75 55 To rque (i n-lb) 16,643 12,000 Seismic Acceleration NS (g) 7.0 3.0 EW (g) 7.0 3.0 Vertical (g) 7.0 3.0 Operator Weight (lb) 350 330 Center of Gravi ty X (in) 10 4.33 Y (in) 10 4.d3 Z (in) la 13.5 Frequency to (H2) 59.5 fo > 33 Hz Evaluation against ASN Section III ASME Section II' Jesign 4 Level A design 4 Level A Table 2 Val ve Numbers 8-inch HBE-M0-1001, 8-inch HBE-MO-10012 Model data for aerodynamic torque modi fication: mi tered el bow two di xneters upstream.
Geometry 2 3.
Al l to rq ue s i n i n-l bs.
(Positive torques tend to close valve)
Model Test Actual Torque for Turque Torque for Val ve Val ve Straight Flow Modi fica tion Installed Candition Angle Angle Normal
- Maximum **
Factore Normal
- Maximum **
80 90 0
2,432 1.2 0
2.914 70 80 0
4,396 1.2 0
3,llo 60 70 0
S,313 1.2 0
6,375 50 60 0
S,4S7 1
0 3,437 40 50 0
6,088 1
0 S.088 30 40 0
4,490 1
0 4,400 20 30 0
3,958 1
0 3.458 10 20 0
3,779 1
0 3,770
- For 140.FM.
- At SS psid.
- Based on test data.
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. Table 3 Electric Actuator Torque Characteristics Torque Torq ue (in-lo)
Switch Torque Torque to Fail Valve Design Trip Switch Appl i ed
- Weakest **
Size To rque Actuator Torque Setting to Valve Component Safety (Inch)
(in-lb)
Model (in-lb)
Required (in-lb)
(Key)
Factor
- 8 10,000 SMS 6,360 1.0 10,800 37,970 3.S IS-H2BC
- At normal voltage 460 ac (higher than design or torque switch trip due to inertia).
- Based on typical key mechanical properties.
Table 4 Electric Actuated Valve Torques (Torques in-lb)
Valve Valve Nominal Val ve Size Numoer Seatina Torque 1
2 3
4 8
HB E -B F-M0- 10006
-10006
-10007
-10008 10,000 6,375 37,970 28,603
- 10. h10
-10009
-10010
-10011
-10012 1 Maxinum aerodynamic torque (L'sitive torques tend to c'ese valvei.
2 Torque to yield actuator key.
3 Actuator safe structural torque.
4 Actual actuator o;tput torque.
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the Clow 8-incn wafer stop valve. Table 1 represents this comparisan and dtwonstrates tnat the loading and dimensional parameters are adequately en-veloped by the generic nuclear qualification parameters. Seismic qualifica-tion has been addressed by a conbination of tests and analyses.
The statt concludes that the subject valve assemblies for PGE fully meet the structural.
seismic, and operational requi rements of speci fication TM-026.
Rev. 1. and ASME Section III 1980 Edition through Winter 1982.
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. 5.0 Summary We have completed our review of information submitted to date concerning operability of the 8 (3-inch) Clow containnent purge and vent valves at the Trojan Nuclear Plant. We find that the information submitted has satis-f actorily demonstrated the ability of the </alves to close against the buildup of containnent pressure in the event of a DBA/LOCA.
Principal Contributors:
R. Hodor, Bf1L R. Wright, NRR i
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s, GENERIC EVALUATION OF THE BADI0 LOGICAL CONSEQUENCES OF ACCIDENTS WHILE PURGING OR YENTING AT POWER MULTI-PLANT ACTION ITEM 8-24 The release of radioactivity through vent or purge valves from a potential large LO' A at power has been considered generically to assure C
that such events do not constitute an undue hazard to the people residing around operating reactor sites.
To evaluate the radiological consequences of such accidents, the following assumptions have been nade:
a.
vent and purge valve isolation signals, circuitry and purge valve actuation are reliable; b.
purge systen isolation valve closure times are generally sufficient to prevent the release of activity associated with fuel failures that could follow a large break (a total accident elapsed time of about 15 seconds or less);
c.
maxinun allowable coolant iodine equilibrium and spiking activity linits do not exceed those contained in Standard Technical Specifications (STS);
d.
fission products generated by pipe breaks are reflective of coolant activity and fuel failures estinated using 10 CFR Part 50, Appendix K, analysis techniques; and e.
radiological consequences of accidents while purging or venting would be bounded by those produced by a large break.
A large number of staff evaluations of the radiological consequences of LOCA's have been performed for construction pennit, operating license, operating ifcense amendment, and Systematic Evaluation Program reviews.
In addition, a generic assessment of the amount of radioactivity that could be released while venting and purging from a spectrum of pipe breaks through the range of purge valve sizes utilized by industry has been made.
In virtually all cases, the contribution through vent or purge valves is esticated to be of the order of 2 percent, or less, of the Exclusion Area Boundary (EAB) and outer boundary of the Low Popula-tion Zone (LPZ) doses that would occur from a large break LOCA in which a source term indicative of a substantial melt of the ccre with subse-quent release of appreciable quantities of fission products is assumed.*
For dose assessments in which only activity in primary coolant systems would be released, or for events in which fuel failures indicative of 10 CFR Part 50, Appendix K, LOCA analyses are indicated, EAB and LPZ dose
, estimates are substantially less than dose estimates made for a large break LOCA assuming a substantial fuel melt.
Since the nagnitude of the vent or purge contribution to severe LOCA dose estimates is small compared to other LOCA scenarios within design bases, we conclude that the consequences of such accidents are within applicable dose guidelines.
A generic assessr:ent of the radiological consequences of large break accidents, including a resulting severe LOCA of the type hypothesized l
for site suitability purposes, while venting or purging at power i
indicates that the dose centribution through open valves is small.
l Therefore, we find total accident radiological consequences of such accidents would be less than the dose guidelines of 10 CFR Part 100
- Estimates based upon SRP analysis techniques and 10 CFR Part 100.11.
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