ML20235F528
| ML20235F528 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 07/02/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20234C612 | List: |
| References | |
| TAC-63576, NUDOCS 8707130428 | |
| Download: ML20235F528 (5) | |
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1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l
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RELATING TO SAFE EMERGENCY SHUTOOWNS~(STEAM GENERATOR MATERIALS) l 1
PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION
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l LICENSE OPR-34, DOCKET NO. 50-267
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1.0 INTRODUCTION
By letter P-87002 dated January 15, 1987, the Public Service Company of I
Colorado (the licensee) proposed to eliminate reliance on only the reheater section of the steam generator for safe shutdown cooling at the Fort St.
Vrain Nuclear Generating Station. The change is necessary to support safe l
shutdown cooling from power levels above thirty-nine percent.
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The licensee performed analyses confirming that one reheater section was not adequate to provide firewater safe shutdown cooling after a 1 -hour inter-l l
ruption of forced cooling (10FC) with the Fort St. Vrain reactor operating at 105% power. The fuel temperature was estimated to peak at 3024 F from j
this power level, exceeding the 2900 F temperature limit established in the Final Safety Analysis Report (FSAR).
Further analyses indicated that safe shutdown cooling could be performed with one EES section supplied with firewater after a 1\\-hour 10FC from power levels up to 87.5% without exceeding a fuel temperature of 2900 F.
I The purpose of this Safety Evaluation is to assess the capability of the steam generator to maintain structural integrity for firewater safe shutdown cooling from 87.5% reactor poe r utilizing the EES section from one steam generator j
following 1\\ hour IOFC.
An evaluation of this transient was reported in General Atomics (GA) Locument 909269 N/C, attached to letter P-86683, Public Service Company of Colorado, dated December 30, 1986.
The primary system pressure and the hot module inlet helium temperature were calculated as shown in Figures 4-4a and 4-5a, respectively.
The primary pressure was shown to decrease to about 600 psia during the initial phase of 10FC, abruptly increasing to about 640 psia on initiation of circulation, then abruptly decreasing to about 350 psia for the remainder of the 10-hour period. The helium inlet temperature was calculated to increase to'about 1300 F during the flow l
interruption and further increase to 1500 F after the start of forced flow, l
then gradually decrease to about 300*F during the next 8-hour period.
The temperature was calculated to remain above 1400 F for a 1\\-hour period.
- 2. 0 EVALUATION The steam generator modules were designed, fabricated and inspected to the requirements of the ASME Boiler and Pressure Vessel Code, 1965 Edition, B707130420 870702 PDR ADDCK 05000267 P
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- including Winter 1966 Addenda, and the Standard Code for Power Piping, USAS B31.1, 1967 Edition.
The ASME Code was supplemented by the following Code Cases: 1319, 1325-3, 1330-1, 1331-4, 1342-1, 1351, 1352, 1361, 1362-1, 1383 and 1389.
The layout for the steam generator modules consists of five sections of helically fabricated tubing.
Referring to GA Document 909190A, the reheater and superheater II were constructed with 1 in. 0.D. by 0.205-in. wall Sanicro-31 tubing; the superheater I and evaporator were constructed with 1-in. 0.D. by 0.125-in. wall 2k Cr - 1 Mo alloy; and the economizer was constructed with 1-in. 0.D by 0.187-in. wall % Cr-% Mo alloy tubing. The feedwater inlet was constructed with 1.25-in. 0.D. by 0.187-inch wall plain carbon steel tubing.
GA Document 909190A records the result of a structural evaluation of the most critical regions of the steam generator during a single cycle cooldown from power using firewater in one reheater module.
It was shown that this one time event would not violate the integrity of the pressure boundary, provided the local helium temperature was limited to 1350"F maximum and not remain above 1300 F for more than one hour.
Creep buckling collapse of the steam generator tubes was identified as the principal failure mode and the primary stress loading for this event.
Specific phonomena of concern were the surface strains caused by thermal shock in the cold-worked material of the reheater tube at small-radius bends, and the plastic strain accumulation in the tube bundle due to local yielding and elastic follow-up.
l The creep buckling computer program (BUCKLE) was used to calculate tube collapse time; that is, the time for the tube ovality to attain the value which caused the maximum local stress to equal the material yield stress. The program uses constant values for temperature and external pressure. The study showed that the most critical steam generator tube for creep buckling above 1330 F was the economizer at the feedwater inlet. 'At temperatures below 1330'F, the carbon steel feedwater inlet tube was identified as the most critical material.
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The results of a structural evaluation of the critical regions of the steam
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generator during a single cycle cooldown from 105% power using firewater in the EES tube bundles were presented in GA Document 909204 N/C. The local helium temperature and pressure varied during the event:
for the first 1% hour period, the temperature was less than 1300'F and the pressure peaked at 700 psig; for approximately 2-hours following the initial period, the temperature peaked at 1660'F and the pressure decreased to 350 psig.
The structural evaluation showed that the reheater tubes would not undergo creep buckling collapse failure during a single firewater cooldown event provided that the reactor power level was limited so that the maximum helium temperature was less than 1660 F for a period less than 20-hours, and the maximum helium pressure i
, l was not more than 350 psig.
Further, the evaluation showed that creep buckling collapse would not be a viable mode of failure at 1500 F.
The calculated collapse time in hours for reheater tubes with 12% cold-worked ovality were calculated at 196 hours0.00227 days <br />0.0544 hours <br />3.240741e-4 weeks <br />7.4578e-5 months <br /> for a temperature of 1500 F and pressure of 822 psig.
After reviewing the analyses presented on the capability of the steam generator to withstand the firewater cooldown transient following a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 10FC, the staff (and consultants) expressed specific structural and metal-lurgical concerns to the licensee.
A summary of the licensee's response j
is presented below:
a)
The structural effect of introducing firewater into the hot steam generator was addressed.
A review of the original design calculations indicated compliance with ASME mechanical limits and fatigue usage factors.
The introduction of cold water would result in a strain of 0.49% compared to an estimated end-of-life ductility of 8% for Sanicro-31.
b)
The potential for either vapor-lock or water-hammer was addressed in to letter P-86682.
After reviewing the geometry, design, heat transfer and hydraulic characteristics of the EES section, it was concluded that with one firewater pump operating and the primary coolant flow rate controlled, flow could be established and maintained in all modules without either vapor-lock or water-hammer (flow stoppage) in any of the modules.
I c)
The concern was raised that a rapid increase in system pressure could occur if steam bubbles were to form in a steam generator tube during cooldown, resulting from the expansion of flashing water.
Emergency procedures preclude this condition. Once the steam generator is flooded, helium flow is controlled to maintain adequately subcooled liquid at the '
steam generator outlet.
d)
Concern was expressed on the metallurgical effect of prior service on l
properties of the steam generator tubes.
Sanicro-31 (Alloy 800H) tubes are potentially subject to metallurgical change, including recrystallize-tion, sensitization, and precipitation aging, which may have a negative influence on the fatigue life, fracture toughness, and ductility.
The loss of fatigue life is not a concern as the transient would be single-cycle event resulting in plant shutdown.
Extensive long term experience with the material in welded construction in elevated temperature service, together with considerable laboratory data, indicate that Sanicro-31 retains sufficient toughness to assure notch-ductile behavior under I
the loading rates applicable to the cooldown transients.
The reduction in residual ductility, however, raised concern about the tubes ability to service the thermal shock and increased bending loads caused by the l
cold 1irewater flowing through the hot tubes.
Reports were submitted showing that the tubes and their welds have ample margin to survive the cooldown transient at any point in the plant's life.
Additionally, sensitization to stress corrosion cracking does not present a problem due to the short duration of the transients.
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l The second group of materials (2 Cr-1Mo, Cr-hMo, and plain carbon steel) may be subject to corrosion when exposed to the chemically untreated firewater for extended periods.
The corrosion rate could be enhanced by erosion caused by particulate found in the fire water.
For the low flow velocities involved and for the length of the cooldown, corrosion of these tubes is not expected to be significant.
The licensee has reported two steam generator leaks.
The first leak occurred in November 1977, in Loop 1 and the second occurred in December 1982, in Loop 2.
The leaks were estimated to be 0.003 inch in diameter.
They were located in the Sanicro-31 material in the superheater II section of the steam generator.
Sections of the Sanicro-31 tubes were removed for metallurgical examination during the plugging operation.
The Fe-Cr-Ni oxide on the surface had an average thickness of 0.008 inch with no evidence of pitting, cracking or corrosion / erosion damage.
The metallographic structures of the Sanicro-31 alloy was considered to be typical as received, fine grained with evidence of some cold-work, free from carbide precipitation.
The material showed no sign of significant metallurgical degradation from service.
The licensee concluded l
that the cause of leakage could not be determined and was random in occurrence.
The properties of cold-worked and recrystallized Sanicro-31 (Alloy 800H) were reported in Vo'.umes I and II of EPRI-HTGR 86-03, " Properties of Recrystallized Alloy 800H and Associated HTGR Steam Generator Design Implications," July 1986.
It was found that at 1350 F, 20% cold-worked material would recrystallize to about 35% by volume in a design life of 30 years.
Recrystallization would lower the rupture strength and increase the creep crack growth rate for a given applied stress and temperature.
The kinetics of the recrystallization process and the effect on the metallurgical properties of Sanicro-31 were calculated for the Fort St. Vrain steam generator in Volume II of the referenced report.
- 3. 0 CONCLUSIONS We conclude from our review of the materials of construction and the analyses submitted by the licensee that the Fort St. Vrain steam generator is capable of maintaining structural integrity during a firewater safe shutdown cooling transient from 87.5% reactor power using the evaporator-economizer-superheater (EES) section for cooling following an 1 -hour interruption of forced cooling (IOFC).
The conclusion in part is based on the analyses of the transient showing a maximum inlet helium temperature of 1500 F at a pressure of 350 psia.
The temperature was estimated to remain above 1400 F for a period of 1\\-hours during this single event. This transient condition represents the worst case conditions for an emergency plant shutdown. Under almost all other conditions, when high pressure feedwater is available, both the reheater and EES portions of the steam generator are flooded.
This would cause the materials in the steam generator to be adequately cooled and, not subjected to high temperatures.
In this analysis, the demonstration of material and structural integrity assures that the steam generator can remove heat from the reactor core
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I under a severe transient.
Thus, under severe transient conditions, the l
fuel temperature can be maintained under accepted temperature limits, and l
the plant can be safely shutdown.
Principal Contributor:
F. Litton, METB Dated: July 2,1987 l
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