ML20037D371
| ML20037D371 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 06/29/1981 |
| From: | Linder F DAIRYLAND POWER COOPERATIVE |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| TASK-03-07.D, TASK-3-7.D, TASK-RR LAC-7634, NUDOCS 8107100147 | |
| Download: ML20037D371 (10) | |
Text
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D DA/RYLAND h
[=7 OOPERAT/VE. eo box EU. 615 E AST AV SOUTH. t A CROSSE WSCONSIN $4601 (608) 788 4 000 June 29, 1981 In reply, please refer to LAC-763ti DOCKET NO. 50 t109 U. 3. Nuclear Regulatory Commission ATTN:
Mr. Darrell G. Eisenhut, Director Division of Licensing PMI$j Office of Nuclear Reactor Regulation s
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Washington, D. C.
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SUBJECT:
DAIRYLAND POWER COOPERATIVE JUL 0 01981 LA CROSSE BOILING WATER REACTOR (LACBWR) F u.s. wcuaa 'N'*"
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PROVISIONAL OPERATING LICENSE NO. DPR il5 s
NT, NT N STbTURALINTEGRITYTESTS
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REFERENCE:
(1)
DPC Lettar, LAC-7387, Linder to Eisenhut, dated February 27, 1981.
Gentlemen:
Enclosed find Safety Evaluation Report (SER) for Containment Structural Integrity Tests (SEP-III-7.D) tihich we have prepared for the La Crosse Boiling Water Reactor.
Our letter, Reference 1, identified topics for DPC to submit for NRC evaluation.
The subject topics were listed in the schedule submitted with Reference 1.
If there are any questions regarding this letter, please contact us.
Very truly yours, DAIRYLAND POWER COOPERATIVE Frank Linder, General Manager FL:llT:ee cc:
J. G. Keppler, Reg. Dir., NRC-DRO III O
NRC Resident Inspectors y
/ /
8107100147 810629 PDR ADOCK 05000409 P
s LA CROSSE BOILING h'ATER REACTOR SYSTEMATIC EVALUATION PROGRAM SAFETY EVALUATION REPORT TOPIC III.7.D CONTAINMENT STRUCTURAL INTEGRITY TESTS
_ Introduction In order to assure that = steel containment structure will respond satis-factorily to the postulated design pressure loads, a program of measure-ments, under the Containment Structural Integrity Test, is required to demonstrate the adequacy of the structure with respect to the quality of construction and material.
The scope cf this safety topic evaluation is to review the adequacy of the structural integrity testing procedures used by the licensee and, using current review criteria as a basis, to evaluate the,_ measurements taken during the testing.
Current Review Criteria The current review criteria for this specific safety topic are:
1.
Standard Review Plan, Section 3.8.2; 2
Regulatory Guide; 1.57 3.
ASME BPV Code Section III, Subsection NE-6000 Evaluation Description of Containment Structure The containment building is a right circular cylinder with a hemis-pherical dome and semi-ellipsoidal bottom.
It has an overall internal height of 144 feet and an inside diameter of 60 feet, and it extends 26 feet 6 inches below grade level.
The steel shell thickness is 1.16 inches, except for the upper hemispherical dome, which is 0.60 inch thick.
The containment building is designed to withstand the instantaneous release of all the energy of the primary system to the containment at-mosphere at an initial temperature of 80 T., neglecting the heat losses 0
from the building and heat absorption by internal structures.
I The interior of the shell is lined with a 9 inch thick layer of concrete, to an elevation of 727 feet 10 inches, to limit direct radiation doses in the event of fission-product release within the containment building.
The containment building is supported on a foundation consisting of concrete-steel piles and a pile capping of concrete approximately 3 ft thick.
This support runs from the bottam of the semi-cllipsoidal head at about el 612 ft 4 in. to an elevation of 621 ft 6 inches.
The 232 piles that support the containment structure are driven deep enough to support over 50 tons per pile. (Reference 1)
The contair. ment bottom head above el 621 ft 6 in, and the shell cylinder from the bottom head to approximately 9 in, above grade elevation (639 ft 9 in.) are enveloped by reinforced concrete laid over a b in, thickness of premolded expansion joint filler.
The reinforced concrete censists of a lower ring, mating with the pile capping concrete.
The ring is approxi-mately 4h ft thick at its bottom and 28 ft thick at a point 11 ft below its top (owing to inner surface concavity).
The ring then tapers extern-ally'to a thickness of 9 in, at the top (el 627 ft 6 in.), and the 9 in.
thickness of concrete extends up thr wall of the shell cylinder to el 639 ft 9 in.
The filler and concrete are not used, however, where cavities containing piping and process equipment are immediately adjacent to the shell.
Except for areas of the shell adjacent to other enclosures, the exterior surface of the shell above el 639 ft 9 in. is covered with lb in, thick siliceous filter insulation, faced with ali1m The insulation of the dome Is Johns-Manville Spintex of 9 lb/f t.3. inum.
density, faced with embossed aluminum sheet approximately 0.032 in. thick.
The i vertical walls is Johns-Manville Spintex of 6 lb/ft,gsulation of the density, faced with corrugated embossed aluminum sheet approximately 0.016 in, thick.
The insulation minimizes heat losses from the building and maintains the required metal temperature during cold weather, and reduces the summer air-conditioning load.
The shell includes two airlocks.
The principal access to the shell will be through the personnel airlock that connects the containment building to the turbine building.
The airlock is 21 ft 6 in. long between its two doors, which are 5 ft 6 in. by 7 ft and are large enough to permit passage of a spent fuel element shipping cask.
The containment building can also be evacuated, if necessary, through the emergency airlock, which is 7 ft long and 5 ft in diameter, with two circular doors of 32b in.
diameter (with a 30-in. opening).
Both airlocks are at el 642 ft 9 in.
and lead to platform structures from which descent to grade level can be made.
When the doors are closed, a clamp exerts a positive force, which is transmitted through the doors to live-rubber gaskets around the door frames to ensure gas rightness.
The airlocks and doors are designed to remain gastight under a pressure of 52 psig from inside the containment shcIl and 1 psig from outside of the shel].
The airlock doors are manually operated.
. t
An 8 ft-by-10 ft freight door opening in the containment building accomo-dates large pieces of equipment.
It will be used only when the reactor is shut down and only if large pie ces of equipment must be removed.
During operation, 9-in.-thick concrete blocks are placed on the outside of the door for shielding.
The door is bolted internally to the door frame in the shell.
Two rubber gaskets in parallel between the door and door
' frame ensure a pressure-tight seal.
Approximately 300 MI cables and 75 bulkhead conductors penetrate the con-tainment shell.
Two 30-ton air-conditioning units keep the containment building air temperature at or below 80 P.
The units also have steam heating coils to provide heat to the building, in case a prolonged shutdown is necessary during the winter, and provide air circulation throughout the building.
Air is exhausted from the building by means of an exhaust blower, which discharges approximately 5000 cfm through a pair of isolation dampers and through the containment shell penetration into the suction plenum side of the stack blowers.
The exhaust air is pulled through two filters located just upstream of the exhaust blower.
The first filter is a coarse filter to remove large particles, and the last is a high-efficiency type filter. An automatic bypass permits the exhaust blower to continue to circulate air whenever the containment isolation dampers are closed.
Ventilation air is admitted into the building through the containment penetration a~d a pair of isolation dampers, and then through duct work to the suction side of the blower of each air-conditioning unit.
Excessive external pressure on the building is prevented by two vacuum breakers which start to open when the negative pressure within the build-ing exceeds 0.2 psig.
A 42,000-gal storage tank in the dome of the containment building supplies water for the emergency core spray system and the building spray system.
A 50-ton traveling bridge crane with a 5-ton auxiliary hoist is located in the upper part of the containment building.
The bridge completely spans the building and travels on circular tracks supported by a ring of concrete around the inside of the building just below the hemispherical upper head.
The crane is operated from a pushbutton station suspended from the trolley.
DESIGN DATA AND STRUCTURAL DESIGN General The design and construction of the aa ntainment vessel conforms to the applicable requirements of the 1962 edition of the ASMC Boiler and Pressure Vessel Code,Section VIII, Unfired Vessels, and applicable code cases 1270N, 1271N and 1272N.
The containment vessel has the ASFE code stamp, i
( l
Design Conditions for Code Calculations The design conditions for the code calculations are as follows:
1 (1) maximum internal pressure 52 psig (2) maximum negative pressure...
0.5 psig i
(3) maximum temperature..
28 Dor (4) minimum temperature........'......
-200F l
(5) welded joint efficiency.............
1024 2
(6)
Basic wind pressure.
20 lb/ft Materials Forgings conform to ASTM A350, LF1, and piping conforms to ASIN A333.
Containment vessel plates and reinforcements and their welded attach-ments, conform to the applicable requirements of ASTM A201 (" Standard Specification for Carbon-Silicon Steel Plates of Intermediate Tensile
]
Ranges for Fusion-Welded Boilers and Other Pressure Vessels") for Grade B steel.
Plates also meet the applicable test requirements for ASTM A300 (" Standard Specifications for Steel Plates for Pressure Vessels j
for Service at Low Temperatures").
Drop weight -tests on the 31 heats of ASTM A Grade B materials used in the containment vessel indicate a nil ductility tempercture range of j
-30 E to -60 f.
Thermocouples are provided on the vessel external surface at locations given in Table C-1 in order to ensure that the shell temperature is always above 0*F (NCT + 30).
Construction The containment vessel is airtight, with a maximum leakage of less than 0.1 percent of the contained ' volume per 24 hr at design pressure (52 l
psig).
The out-of-roundness does not exceed 0.5 percent.
I Radiography i
l All butt welds were fully radiographed in accordance with Paragraph UW-51, ASME Boiler and Pressure Vessel Code.
All welds of doors, nozzles, and opening frames, and all welds that could not be radio-
- graphed, examined for cracks by the magnetic-particle or fluid-
]
penetrant mthods of inspection.
Testing Requirements for the initial testing of the containment vessel included the following:
l l
l (1) visual inspection of welding (2) pressurizing the containment vessel to 5.0 psig and perform-ing soap bubble tests on all welds (3) pressurizing the shell to 59.80 psig, then reducing the pressure psig for fina) soap bubble tests on all welds at a vessel pressure of 50 psig.
The 52 psig pressure was maintained for a 3-day period, with leakage measurements taken hourly.
The inner chamber method was used to measure leakage.
After completion of the vessel penetrations, a final leak-rate test was performed at 52 psig for two days (preoperational leak-rate test).
Vacuum Breakers Vacuum breakers are set to start opening at a 0.2 psig-vacutm and to be fully open at a 0.5-psig vacuum.
They will remain tight at the 52 psig design pressure without damage to the seat, gasket, diaphragm, or casing.
Each vacuum breaker is mounted inside a test chamber, which permits per-Iodic testing of the vacuum breaker for leak-tightness.
During normal operation, the upper half of the test chamber is jacked up from the lower half to permit air to flow into the vessel whenever the negative pressure exceeds 0.2 psig.
Contained Energv The containment building can contcin all the steam and water released from the primary system in the tvent of a major system rupture.
The free-air volume of the containment building is 204,160 ft3 If all the energy of the primary system is assumed to be instantaneously re-leased to the containment atmosphere at an initial temperature of 80 F, and if there-ere no heat losses from the building or heat absorption by internal structures, the resulting pr ssure and temperature would be 48.5 psig and 273 F.
These values compare favorably to the design values of 52 psig and 280 F.
LEAK TEST PROCEDURES AND ASSES 3 MENT OF RESULTS The overload test procedure fulfilled the requirements of the ASME Code,Section VIII, as modified by Code Case 1272 N.
the method used for the leakage test consisted basically of comparing the pressure in the conta inment vessel with that in an airtight inner chamber.
The inner chamber was proved tight by thorough preliminary inspection methods, and any relative decrease in containment vessel pressure after temperature equilibrium was regarded as from external leakage.
A high degree of sensitivity to this pressure differential was achieved by use of a water manometer to measure the pressure differential betwean the two air volumes.. _ - - _ _ - _ - _ _ _ _ _ _ _ _ _ _
- _ _ _ _ _ _ -. ~
o I
Prelfminary Check 1
Preliminary testing was performed in the shop and field before the over-load and leakage-rate tests at the site.
All airlocks were shop tested for tightness and for operation of door mechanism, including the equal-izing valves, and were 'ound to be adequate.
All shop-welded manholes and nozzles were magnaf:.uxed inside and outside after shop stress relief; there was no indication of cracks or other defects.
Overload Test j
follow?ng construction the overload test was performed by the constructor (Reference 2).
The vessel was pressurized to 5 psig and all vessel connections and welds were checked by a soapsuds inspection.
The air i
pressure was then increased in increments to the test pressure of 59.8 psig. After a 1-hr holding period, vessel pressure was reduced to the design 52 psig.
The se':ond and final soapsuds inspection was conducted at design pressure.
The outer. door of each lock was left open.
The soapsuds inspection at 5 psig found several minor leaks in connections but no leaks in welded seams or in gaskets.
All leaks were repaired and were then found to be tight under soapsuds inspection.
The overload test I
to 59.8 psig on the vessel and the airlocks was successfully con.pleted.
The final soapsude inspection at 52 psig found no leaks in welded seams.
Initial Leakage Rate Test Following the overload test, the constructor commenced the initial leakage
[
j rate test.
The pressurn was adjusted so that maximum containment pressure would be approximately 52 psig during the holding period for the leakage rate test.
Water was introduced into the differential manometer to approx-Imately the mid-height of the scale. Air was then pumped into the contain-ment vessel until the water manometer indicated a pressure about 10 inches higher than that in the inner chamber system.
The pressure and temperature readings were recorded hourly for four i
successive nights and the average leakage per 24-hr period is calculated to be:
% leakage
= 1/3 x x 100 0.11 0.002%
=
2u-hr 129.80 x 13.6
(
The calculated leakage is well within the specification value of 0.1 percent per 24-hr period.
The measured tightness of the vessel is consistent with the results of the soapsuds inspection.
Pre-Operational Leakage Rate Test After installation of all building penetrations, the preoperational leakage rate test was performed by Allis Chalmers (Reference 3).
Individual penetrations were tested to establish their leak rate.
The recator building leak rate was again measured, using the same reference vessel method.
The test was performed similarly to the initial leakage rate test.
The containment building leakage rate l
1 i
.__,.___m_.___,_____.._
,,__.__._....,,,,______._.m_._,_...__
was found to be 0.038 percent of contained volume per 2il h as measured over a two-day interval.
- r period limit.
A detailed description of the structure integrity test for thThis is w La Crosse Boiling Water Reactor Containment Building is Reference 2, e
Pre-operational test information is contained in Refere contained in 3
nce In-Service Leakace Tests There have been eight in-service integrated leakage rat on the La Crosse Boiling Water Reactor Containment Buildin through 10).
e tests performed The last two were done using the mass balance methodThe first 6 w g
vessel method.
Based on the results from all the in-service tests it i The tests results also show that the integrity of theno deter s apparent that as occurred.
penetrations is acceptable.
containment vessel Comparison to Current Revicw Criteria Boiling Water Reactor Containment Vessel conformed tT s
on the La Crosse of Article NE-6000 of Reference 11 o all the requirements The first two in-service leakage rate tests were ance criteria was more stringent than present day acc The accep t-0.055Wday of contained volume vs. 0.0757/ day ance values e.g.
leakage rate tests were conducted by the internal refThe next four in-service per the requirements of Reference 12 with acceptance erence vessel method by LACBWR Technical Specifications as < 0 OR{/ day criteria established less than the present day acceptance criteria of contained volume, The leakage rate tests of Reference 7 and 8 were l
50, Appendix J. cable requirements of Title 10 of the Code of Pederal Rconducte with the acceptance criteria of 10 CPR 50 Appendix JLACBWR Tec still used the internal reference vessel method o comply The test method mined inaccordance with Reference 13 Leakage rates were deter-The leakage rate tests of Referenets 9 and 10 wer specifications, which meet the acceptance criteria of 1 e conducted to the Technical J.
The test method used the absolute method of leaka R 50, Appendix specifically, the mass plot analysis technique ge determination, termined inaccordance with Reference 14 Leakage rates were de-Based on our review of the above reports The leakage rates were always within the acceptablot' the c uring any of the tests.
operation after every test.
e limit prior to reactor [N
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Conclusion Eased on the information provided in the test reports and Reference 15 and the evaluation stated above, we conclude that the test procedures used are adequate and the f est results provide a basis to assure that the containment structure will safely perform its intended functions and will continue to withstand the design pressure load of 52 psig.
1 e
f
- O REFERENCES (1)
Sargent & Lundy Report SL-2003, dated l'ebruary 25, 1963.
"Contain-ment vessel Pile Driving Operations for 50 MWe Boiling Water Reactor at Genoa, WI."
(2)
Chicago Bridge & Iron Report, " Final Leakage Rate Determination of Reactor Containment Vessel", dated Mry 1967 (3)
ACNF-68511 La Crosse Boiling Water Reactor Containment Leakage Rate Test, 1968.
(4)
LACBWR Containment Vessel Class-A Integrated Leak Rate Test of October 1969, DPC-851-27.
(5)
IACBWR Containment Vessel Class-A Integrated Leak Rate Test of November 1970, DPC-851-34 (6)
Containment Building Leak Rate Test, September 1971, DPC-851-35.
(7) 1975 Containment Building Integrated Leak Rate Test, LAC-TR-032, September 1975.
(8) 1978 Reactor Containment Building Integrated Leak Rate Test, LAC-TR-066, December 1978.
(9) 1979 Reactor Containment Building Integrated Leak Rate Test, LAC-TR-071, May, 1979.
(10) 1980 Reactor Containment Building Integrated Leak Rate Test, LAC-TR-093, December 1980 (ll)
ASME Boiler and Pressure Vesnel Code,Section III, Division I, Subsection NE' " Class MC Components",
American Society of Mechanical Engineers.
(12)
" Proposed Standards for Leakage Rate Testing of Containment Structures for Nuclear Reactors", ANS 7.60, Approved for publication for comme nts by ANS Standards Committee, June 15, 1964 (13)
ANSI N45.4 - 1972, " Leakage-Rate Testing of Containment Structures for Nuclear Reactors."
(14)
ANS-N274, " Containment System Leakage Testing Requirements, " Revision 2, May 15, 1978.
(15)
ACNP - 65544 La Crosse Boiling Water Reactor (LACBhR) Sa feguards Report for Operating Authorization, Revised, August 1967 )