ML20079M717

Speech Entitled, Superheater Tube Leaks in Steam Generators of Fort St Vrain HTGR, Presented at Joint Ans/Asme Jan 1984 Conference

ML20079M717
Person / Time
Site:	Fort Saint Vrain
Issue date:	01/31/1984
From:	Holmes M, Wong A PUBLIC SERVICE CO. OF COLORADO
To:
Shared Package
ML20079M708	List: ML20079M705 ML20079M717
References
TAC-54598, TAC-65991, NUDOCS 8401270400
Download: ML20079M717 (40)

Text

~

)

%-' ATTAcHMmT l To P-84c'le L

t SUPERHEATER TUBE LEAKS in the STEAM GENERATORS of the FORT ST. VRAIN HIGH-TEMPERATURE GAS-COOLED REACTOR JANUARY 1984 i

Paper to be presented at the 1984 Joint ANS/ASME Conference by A.H..Wong and M.H. Holmes NUCLEAR ENGINEERING DIVISION PUBLIC SERVICE COMPANY OF COLORADO P. O. BOX 840

'lENVER, COLORADO 80201 8401270400 840120 PDR ADOCK 05000267 P PDR

I!h/

w. . .

ABSTRACT In December 1982 a tube failure occurred near the bottom of

[

i superheater 2 (SH 2) in module B-2-3 of the Fort St. Vrain (FSV) Loop 2 steam ' generator. This was the recond tube failure at FSV. The first (November 1977) also occurred near the bottom of SH 2 but in module .B-1-1 of the loop 1 steam generator. Since both tube leaks were found at or near a floating tube support plate. Public Service

-Company of Colorado (PSC) and GA Technologies (GA) evaluated potential causes for failure at this location. Both have concluded that_ the two leaks were purely random failures of the Alloy 800 tube material and cannot be explained by the available evidence.

t

, . ,e . -e ...-.g,,- . - - ,-. ,--.-n ---w- , ,. , - - . n.

s

.g_

BACKGROUND

[ FORT ST. VRAIN REACTOR DESCRIPTION

.The Fort St. Vrain (FSV) high temperature gas-cooled reactor It is a helium cooled (HTGR) is designed to generate a net 330MWe.

and graphite moderated rentor having two identical once-through steam generators and four steam-driven helium circulators. The steam generators and helium circulators are contained in the prestressed concrete reactor vessel (PCRV) cavity located below the reactor core.

This arrangement of components in the PCRV reduces the primary coolant inventory and extent of external piping and eliminates the need for steam generator pressure shells and a secondary containment building. Figure 1 is a cut away view of the FSV reactor showing the arrangement of the components within the PCRV. Figure 2 is the simplified process flow diagram of the primary coolant system in the PCRV. Table 1 gives the nominal plant operating conditions for Fort St.-Vrain.

FORT ST. VRAIN STEAM GENERATOR DESCRIPTION Each of the two identical steam generators consist of six modules. The twelve modules are spaced equally apart in a circular l

array in the PCRV. Six adjacent modules collectively are the loop 1

( steam generator and the other six makeup the loop 2 steam generator.

! A typical module is shown in Figure 3. _

l' Hot (690 psia, 1427 degrees F) reactor coolant helium gas from 1

the reactor outlet plenum enters the individual modules through a top duct, and constrained by the outer shroud, flows downward over the I

L

5 modules water and steam tubes. The helium passes successively over i

the reheater, superheater 2 (SH 2), superheater 1 (SH 1), evaporator and economizer. The economizer, evaporator and superheater 1 (EES) constitute an integral tube bur.dle. The cold (686 psia, 741 degrees F) helium exits through an opening in the shroud and enters the suction of the two helium circulators associated with the module's loop. Figure 3A is a photograph of the assembled EES bundle with the SH 2 bundle shown on the left side of the photograph.

Feedwater (3102 psia, 413 degrees F) from below the PCRV enters the feedwater ring header of each module where trim valves are used to distribute the ' rater to eighteen subheaders which first penetrate the secondary and then the primary closure of the PCRV. Inside the PCRV, each subheader is then divided into three smaller tubes. The fifty-four feedwater tubes then wind helically upwards through the EES where the water becomes steam in the evaporator section of the tube bundle and reaches 750 degrees F at the top of SH 1 before it enters the top of the SH 2 tube bundle. It then winds helically downward through SH 2 reaching 1050 degrees F as it leaves SH 2.

These fifty-four steam tubes are combined into eighteen subheaders before the subheaders penetrate first the primary closure and then the secondary closure beneath the PCRV. The eighteen subheaders then connect into the module's main steam ring header. Main steam leaves the header at 1008 degrees F, 2512 psia.

Figure 4 shows twelve columns (coils) of tubes in the SH 2 bundle with column 1, the innermost column and column 12, the outer most. Three support plates are attached to a central support cylinder with a limited flexibility connection. As shown in Figure 4, the plates are equally spaced around the tube bundle and are

g drilled to accept the helical tube coils. These three plates provide support for the twelve columns of tubes. There is also a floating support plate located between each of the fixed support plates.

These -floating plates support the outer six columns of tubes. A sleeve / wedge' pair is placed on each tube to protect the tube from damage (wear) at each point where the tube passes through a support plate.

In the event of a feedwater or main steam leak within the PCRV, the failed tube (s) cannot be individually plugged. The feedwater and steam subheaders outside the PCRV that serve the failed tube (s) are cut and capped. -isolating -three tubes. A typical capped subheader is shown in Figure 5. Steam generator tube leaks have been analyzed and allowed for in the design and operation of Fort St. Vrain. The

. design of the steam generator allows one subheader in each of the j twelve modules.to be capped without affecting the plant's rated capacity. If the coils are dispersed up to five subheaders in a single module may be capped at full power conditions. Table 2 summarizes the effect of varying combinations of subheader failures

. on reactor coolant helium temperature while maintaining 100 percent

i. load conditions.

Because of the differences between Fort St. Vrain and the typical LWR, Table 3 is provided to show the materials used for the tubes and structures in the FSV steam generators.

The reheater section of the steam generators has not been ,

described because it operates at a lower pressure than the reactor coolant helium in the PCRV. A reheater tube failure does not

' introduce steam (moisture) into the PCRV. Moisture ingress via a i

- - - . . . , , , . , . . m-x..-e .e,m.-...,, .-v....,--.,yr.-_..-- w.-.-_-.m..,%.,,.,,- ,m y. %. m .- - ,~m.m.--,.----,.-..._...-..-.,%---...,,m-.

.. =.

O feedwater or s' .m tube failure is of concern since the moisture causes oxidation of the reactor graphite moderator and internal

components. When moisture in the PCRV exceeds the 10 ppm limit for oxidants in the recctor coolant helium, reactor power level must be reduced to lower the reactor coolant helium temperature below 1200 degrees F for protection of the reactor and the reactor coolant system.

THE STEAM GENERATOR TUBE LEAK At the end of September 1982 while the plant was operating at 70 percent reactor power, a reactor scram occurred as a result of plant protective system surveillance testing. Primary coolant moisture levels began to rise shortly after the scram for no apparent reason.

Based upon previous experience, the increase was assumed to be caused by an unobserved upset in the helium circulator auxiliary system, resulting in water ingress into the PCRV. Based upon the amount of water being removed from the PCRV by the helium purification system over an extended period of time other possible sources of moisture were investigated.

On December 7, 1982, PSC decided that if the source of the high moisture level in the primary coolant helium were traced to a leak in a module of the steam generators, sufficient testing should be done to ascertain the exact location of the leak (s). PSC considered this necessary to assure that any subsequent evaluation of the leak test

~

results might be able to determine a specific cause of the leak.

This is important because the second occurrence of a steam generator l

l tube leak might suggest a trend in increasing numbers of tube i failures similar to those in the steam generators of an LWR.

On December 8, 1982, PSC started testing the EES section in each r of the six steam generator modules in loop 2 in an effort to locate j the module and subheader in which a small feedwater or steam leak existed. Loop 2 was suspected because the radiochemical analysis of the steam dump tank, to which loop 2 steam and water had been dumped, indicated the presence of isotopes from the primary coolant helium.

The leaking subheader was found in steam generator module B-2-3 on December 12, 1982. The inlet subheader is feedwater subheader 13 and the outlet subheader is Steam Subheader N-96. Theleak(s)may, therefore, be in one or more of the subheader's three tubes. These are identified as main steam tube numbers 37, 38 and 39 and are located as shown in Figure 4.

1 s

/

1 t

i EVALUATION OF LEAK LOCATION

[ DETERMINATION OF LEAK SIZE On December 10, 1982, when PSC was 'certain that the high moisture level in the primary coolant helium was due to a steam generator tube leak, the task became one of determining the size of the leak and a means of finding its exact location.

The high moisture level in the reactor coolant helium required the plant to be shutdown and depressurized for finding the source of moisture in-leakage. The leak rate was assumed to be about 5 gallons per day based upon observation of the moisture removal rate attained by the helium purification dryers while maintaining a relatively constant moisture level in the primary coolant helium with a 500 psi differential between the steam side of the steam generator tubes and the helium in the PCRV. Based upon the previously observed 5 gpd leak rate and 500 psi differential pressure for the moisture (water) leak from the feedwater/ steam tube into the primary coolant, the leak hole size was calculated using the Darcy equation for flow of water through an orifice. It was estimated to be about equivalent to a 0.003 inch diameter square edged orifice. This leakage rate was corroborated during the testing that found the location of the leak.

With this very small leak, it was believed that the rate of gas pressure decay due to the flow of gas through the leak hole would be significantly greater than that when water is flowing through the _

same hole with the same pressure drop. It was also believed that the hole (leak) elevation could be established by using a water manometer as shown in Figure 6. Calculations confirmed that the leak hole was

small and that the flow rate of argon was much greater than the flow rate of water for the same conditions. Argon was selected as the

, test gas because it is a traceable gas in the primary coolant systtm and its increasing concentration in the steam dump tank is a reliable indication that the leak hole had been identified.

DETERMINATION OF LEAK H0LE ELEVATION Since the November 1977 tube leak was determined to be in the bottom (outlet) end of superheater 2 (SH 2) in Main Steam Subheader 9 of module B-1-1, the procedure for locating the current leak was concentrated in this region of SH 2. The leak was found in module B-2-3 at an elevation of 52.4" below the reference elevation level of 4811'-0". This corresponds to an elevation of 4806'-7.6" or 4.1" above the bottom of SH 2.

i The test data curve, Figure 7, which was developed from the water manometer testing, shows a distinct change in rate of the decreasing water level at 57.0" (4806'-3" elevation) below the reference level and 64.5" (4805'-7.5" elevation) below the reference level. This can be explained by the configuration of the SH 2 downcomers, tubes 37 and 38 which have an 8" long horizontal offset

-below the end of the bottom coil of SH 2. See Figure 8. The 4805'-

7.5" elevation is the centerline elevation of the horizontal offset which 'is 8" below the centerline of the bottom coil of SH 2. The 4806'-3" elevation is the point where the bottom inside surface of the bottom coil of SH 2 turns down. See Figure 9. The measured 4306'-3" elevation is therefore considered to be the bottom of SH 2 since it is nearly identical to the elevation of the bottom coil of SH 2 that is shown as 4806'-3 1/2" elevation in the maintanance

4 manual for th2 Steam G:n rator. The 1/2" discrepancy can be resolved by considering two additional dimensions. First, the 4806'-3 1/2" elevation dimension is the center line of the 0.590" ID tube rather than the bottom surface of the tube as shown in Figure 9. This lowers the apparent elevation by 0.295". Secondly, there is the drop to the point where the rate of the dropping water level increases This lowers the apparent elevation by 0.157".

significantly. The addition of these two dimensions is 0.452", which reduces the 1/2" discrepancy to less than a 1/16". Similarly, test data established the top of SH 2 at 4809'-5" elevation, corresponding to the elevation shown in the maintenance manual for the steam generator.

The change in slope (knee) of the Figure 7 test data curve at the 4806'-3" elevation can be seen in Figure 9 as the intersection of the 3 degree slope of the bottom coil of SH 2 with the vertical centerline of the dov";omer. This is the elevation at which there is a significant increase in the incremental water level drop for each equal increment of water drained from the tubes.

.It should be noted that a calculation was also made to check that there is, in fact, no backpressure built-up in the gas leg of the SH 2 tube above the water level at the indicated leak elevation.

This backpressure would depress the water leg and give a higher indication of leak hole elevation on the free side of the manometer.

This corroborates the test technicians who reported no noticeable velocity of gas exiting the gas leg (shown in Figure 7) of the SH 2 tube while conducting the test for finding the elevation of the leak -

hole.

6 After the leak location was established, the feedwater and main steam subheaders were capped, as shown in Figure 5, and the plant was returned to service.

CONCLUSIONS

1. Based upon the leak elevation at 4806'-7.6" and the bottom of the SH 2 at 4806'-3.5", the current leak is 4.1" above the bottom of SH 2 in steam generator module B-2-3.

2. Based upon the leak rate test results, the hole is very small and may be equivalent to a .003" diameter orifice.

3. Because the leak is so small, it may be assumed to be in only one of the three tubes (37, 38 or 39) connected to subheader 13.

i l

l ..

l l

t

- m , , _ . .

e DETERMINATION OF CAUSES FOR THE TUBE LEAK Since the leak in the SH 2 tube (37, 38 or 39) of steam generator module B-2-3 is the se:ond such leak that has occurred in the FSV steam generators, detarmination of its cause, if possible, is desirable to provide information on the trend of the frequency of future tube failures. The major concerns for steam generator tube failures at Fort St. Vrain have been the metallurgy and residual stresses in the reheater's Alloy 800 3D tube bends, weld joint defects, vibration stresses causing fatigue, carburization of Alloy 800, water chemistry, corrosion, wear, cold springing, low cycle fatigue, crack propagation and loss of tube sleeves and wedges.

These concerns have been evaluated and are summarized below:

3D TUBE BENDS As discussed earlier, the tube leak is not in the reheater section of the module but is 4.1" above the bottom of SH 2. Review of the steam generator design drawings shows that the nearest 30 Bends are at the end of the bottom coils of tubes 37, 38 and 39 in SH

2. The leak is 4 to 7 feet up the tube coil from the 30 tube bends.

The metallurgical and residual stress concerns about the 3D Bends are obviated because the leak hole is not in the 3D tube bends.

e WELD JOINT DEFECTS Each SH 2 tube assembly consists of two 35 foot long sections of 1" 0.D. x 0.205" wall Alloy 800 tube material welded together. These are then welded to a downcomer tube. Design drawings show the weld joint between the downcomer tubes and the end coil of the SH 2 tubes to be located 1", 4-3/4" and 8-1/4" from the tube bend for tubes 37, 38 and 39, respectively. The leak location at 4.1" above the bottom of SH 2 is 4 to 7 feet from the end of the coil. The review of the design drawings also showed that there are no instrumentation welds in the SH 2 bundle of module B-2-3.

Since welds are not located near the leak location, weld joint defects are not the cause of the tube leak.

TUBE FIXATION AND CYCLIC OPERATING STRESSES Since the leak hole is in the tube wall, several mechanisms might have accelerated the development of this hole. One of these would be the effect of the tube's fixation at the tube support plates that is shown in Figure 4. This Figure shows a sleeve / wedge pair between the tube and the tube support plate at every point where a tube passes through a tube support plate. If the tubes are fixed in position where they pass through the tube support plates, tube fixation could introduce cyclic operating or startup-shutdown stresses into the tubes where they are fixed to the tube support plates.

~

Table 4 was prepared to determine if there is a correlation between the location of the leak hole and the tube support plates.

I Table' 4 shows that tube 37 intersects floating tube support plate A (FA) shown in Figure 4 at a calculated elevation of 4.14". This is within one-sixteenth inch of the 4.1" leak elevation determined from the leak location testing. FA is the next to last tube support plate that tube 37 passes through before turning down to become a downcomer. Based upon this coincidence, Table 4 was extended to include leaking tube (s), 25, 26 and 27 in module B-1-1 which were plugged in November 1977. The B-1-1 leak elevation was established at 6" above the bottom of SH 2. Table 4 shows tube 27 intersecting plate FA at a calculated elevation of 6.02". This again is within one-sixteenth of an inch of the established leak hole elevation. In this case, FA is the second from the last tube support plate that tube 27 passes through before turning down to become a downcomer.

The coincidence of-the two leaks occurring at or near the same tube support plate (but in two different modules) is intriguing.

Therefore, other similarities were investigated. Both tubes have the same configuration of leadout (downcomer) tube, and both pass through adjacent fixed tube support plate A. However, after passing through Plate A, tube 37 turns down to become a downcomer while tube 27 passes through one more tube support plate, FC, before turning down to become a downcomer.

The steam generator design reports were reviewed to determine if the tube to tu'oe support plate fixation was discussed. While the design reports contained no obvious reference to the effect of operating cyclic conditions or startup and shutdown conditions with ,

the tubes fixed at adjacent tube support plates, the designer of the steam generators did consider the tubes fixed at the tube support

plates and found the condition acceptable for the anticipated design transient conditions.

TUBE VIBRATIONS Since the leak hole in the tube wall has been postulated to be at or near a tube support plate, the mechanism of vibration has been considered because the designer has had some concern for tube vibration due to the ~ axial' (transverse) helium flow over the helically coiled tube bundles. This concern was evaluated in the air flow test that was performed on a spare steam generator module and the hot flow test performed on the instrumented steam generator module, B-1-6, as part of the plant's startup tests. The results of these tests showed no significant tube vibrations. That is, the strain gages mounted on all of the SH 2 tubes were well below the specified limits. The air flow test results showed a measured microstrain of less than 9.0 micro inches / inch. The hot flow test at 28% power showed a microstrain of 7.9 micro inches / inch. Both of these values are well below the designer's specified limit of 30.0 micro inches / inch. The vibration stress calculated at this location is reported to be well below the ASME Code Fatigue Allowable Stress.

The configuration of the downcomer end of tube 41 is identical to tubes 27 and 37. Since the air flow test data for tube 41 showed acceptable strain levels, it can be assumed that tubes 27 and 37 would exhibit the same results.

Flow induced vibration is not considered a contributory cause of the tube leak.

CARBURIZATI0ft 0F At LOY 800 Carburization of the Alloy 800 steam generator tubes has been considered for Fort St. Vrain. Carburization is a process whereby carbon is ' deposited onto metal surfaces and absorbed into the bulk metal to form carbide. Carbide formation could potentially cause embrittlement and weakening of the metal structures. At Fort St.

Vrain, the concern was due to the presence of carbon conoxide and -

methane reactions at temperatures above 1500~F and in a reducing

' atmosphere, i.e., high hydrogen to water ratio. This phenomenon results from gasecus impurities and moisture ingress into the primary coolant helium. The first chec! for carburization at FSV was performed by GA from specimens that were taken from the inlet plenum of steam generator module B-1-5. GA found no evidence of the onset of carburization due to circulating impurities transported by the primary coolant helium from the core to the steam generator.

Review of the tube fixation procedure which was used during assembly of the steam generators revealed the possible use of graphite powder on the sleeve / wedge pairs to aid in assembly of the tube coils with the tube support plates. Immediately following the coiling of each cylinder of tubes into the tube support weldment, this procedure states: " Remove graphite coating on fixation sleeve I

by air blast, using a minimum of 100 psi air pressure." This step in the procedure was followed by the fixation of the sleeve / wedge pairs on the tube between each tube and the tube support plate.

Several persons who witnessed the fabrication of the steam generators recollect that the sleeve / wedge pairs were received with a graphite coating and each pair was wiped clean after installation.

They believe that free graphite could have been left on some of the i

O tubes tihere it may have been trapped between mating parts or was

- inaccessible for removal by the wiping procedure.

The apparent free use of graphite at this point in the steam generator's fabrication suggests the deposition of significantly larger concentrations of a previously unidentified source of carbon (graphite) on the surface of the Alloy 800 type tube material in SH

2. It should also be noted that this graphite is located downstream of the inlet pleunm corrosion specimens and would not be likely to deposit on the specimens because it would have to be transported through the helium circulators, the reactor core and back to the steam generator inlet.

GA reviewed the potential carburization effect on the SH 2 tube ,

material and has stated the amount of graphite used during tube i

bundle assembly and service conditions to date are not probable e

causes for failure of the tubes.

WEAR The concern for wear of the SH 2 tubes by the tube support plates led to the use of a sleeve / wedge pair at every location where tubes could contact another tube or tube support plate. The design of the EES and SH 2 tube bundles lead to many locations where thinning of the tube walls can occur as a result of the tube rubbing against tube support plate. For this reason, a calculation was made to. determine the minimum wall thickness that could result in failure. .

Utilizing simplifying assumptions, a 1.0 inch 0.D. Alloy 800 steam tube operating at 2450 psig and 1300 degrees F has, a 10,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> stress / rupture strength of 11,000 psi per Huntington Alloy data.

~

Based upon the ANSI B31.1 code equation for minimum pipe wall thickness, the calculated thickness is 0.096". Since the nominal wall thickness is 0.205", the tube wearing then must be at least 0.109". With the plant's cycle life to date of about 300 transient events of various types, it is inconceivable that any tube has worn to the point of failure. It is also noted that this condition postulates no protection of the tube by the wedge and sleeve. Tube thinning due to wear is not considered the cause of the leak.

WATER CHEMISTRY - STEAM SIDE CORROSION The effect of the steam generator's water chemistry on the inside surface of the tubes has been considered. Specimens taken from ~the feedwater and steam subheaders connecting with the failed ,

tube in module B-2-3 were evaluated by GA.

GA found the Alloy 800 steam tube to be in excellent condition.

The Fe-Cr-Ni oxide film on the inside surface is uniformly thin and smooth. The tube wall shows no evidence of pitting, cracking, or erosion / corrosion damage. The microstructure is normal for Alloy 800 Grade 1.

GA also investigated fouling of the tube due to water impurities i to detennine if the leak was caused by a change in the thermal loading at the leak region from water side fouling. Fouling is expected to occur in the region of the EES tube bundle where boiling occurs. GA's studies have determined that fouling at FSV has N occurred where predicted and that the predicted ten perature changes at the leak location induced by fouling of the EES are very minor.

O Fouling has a minor effect on the thermal loading at the leak location.

GA found nothing in specimens taken from steam generator module B-2-3 that is different from the results of similar specimens taken from module B-1-1 that were examined by GA following the first (November 1977) tube leak.

There is no signif: cant degradation of the superheater outlet tube due to water chemistry and tube fouling.

CORROSION - HELIUM SIDE The corrosion effect of the high temperature helium on the .

outside surface of the Alloy 800 tubes has been considered.

Frequent moisture ingress incidents during FSV's power range operating conditions have introduced impurities into the primary coolant that could be detrimental to the life of the steam generator tubes. The resulting high moisture levels from past incidents generally remained throughout the subsequent shutdown period. When coupled with air ingress during refueling and other reactor outages or 'with other impurities, high moisture conditions could result in detrimental effects on the helium side of the steam generator tubes.

The metal specimens, one set of specimens in each helium inlet plenum of modules B-1-5 and B-2-5, are used to monitor the effects of primary coolant on various metallic specimens including Alloy 800.

Test results on the first set of specimens that were removed from module B-1-5 following the November 9, 1981 shutdown of the plant

. showed only slight oxidation and no sign of carburization or sulphidization. These specimens had been in place since July 1976.

While metallurgical specimens are exposed to the actual primary coolant helium entering the steam generator modules, the Alloy 800 specimens are only representative of the exposed surfaces of the SH 2 tubes. The specimens, which are smooth cylinderical surfaces exposed to the helium, do not simulate the crevice condition existing between the tube and wedge, the wedge and sleeve or the sleeve and tube support plate where crevice corrcsion could initiate a crack or oxidiation could result in loss of any tube to tube support plate clearanca.

Corrosion in the crevices between the Alloy 800 tubes and tube wedges might be likely if any condensed moisture reaching the crevices contained high levels of dissolved oxygen or contaminants detrimental to Alloy 800. Crevice corrosion is very unlikely because of the chemistry and oxygen control of the water sources that could enter the primary coolant.

e-

~C01.D SPRINGING

. Cold springing during initial steam generator fabrication has been a concern. The concern was the high residual stress 'resulting from cold springing of tubes to make final weld fit-ups during assembly when combined with other phenomena associated with high temperatures and load cycling. This concern was addressed to the 3D bend areas-of.the tubes rather than the straight or helical coil sections of. the tubes. Cold springing is a random effect resulting from fabrication tolerances and procedures.

Cold springing could be considered a cause of the leak if the leak occurred at the last tube support plate before the tube turns down to become a downcomer. However, the first and second tube leaks

-have been postulated to be at or near the second from last and the next to last support plates, respectively.

.The effect of cold springing is not considered a contributor to the failure of the leaking tube.

LOW CYCLE FATIGUE-After -the leak -location was identified, GA re-examined the existing low cycle ~ fatigue damage calculations to determine the critical '. transients and damage expected at the leak location. GA also investigated the previous operating history to determine if the transients experienced by the steam generator were as expected by the 4 design analyses. -

GA's analyses show that the Superheater 2 tubes have a larger l fatigue margin than most of the other portions of the main steam i

- ,,e-, ..- - - _+ , , -,,...,,.,-.------.-.,..,-n. .~,.-.n,--,-,,,,-r - - , , . , , , . , , - . - , . , ~ . . - , . _ , , .

bundle. Th2 most critical region of the Superheater 2 tubes is the inlet (Upper) region due to the stresses caused by the radial thermal gradient associated with the high heat flux. The analysis shows that the lower region of the Superheater 2 bundle where the leak occurred is not a fatigue critical region. FSV records of thermal transients showed that a few events could have been somewhat more severe than those analyzed for the fatigue damage calculations. However, the indicated temperature change rates, although higher, were judged by GA to be not sufficiently high enough to make the lower end of the SH 2 become fatigue critical. The upper region of the SH 2 bundle would still be more fatigue critical under more severe transients than the lower region where the leak occurred.

Since the region of the leak location is not critical in GA's analyses, low cycle fatigue can be dismissed as a cause for the leak.

l

CRACK PROPAGATION Following the determination that a tube failure occurred in the lower end of SH 2, GA evaluated the growth of a small crack that has been postulated to exist at the leak location. Since the stresses existing at the leak location had not been previously calculated, GA evaluated the crack behavior using the bundle stress previously

! calculated for the top of the SH 2 tube bundle at 100% power l conditions because this results in maximum bear hug and tube wall l

l radial thermal gradient stresses.

Using. th'e ASME BPV Code,Section XI, Division 1, Article A-3000 ~

method to obtain "K" values for an assumed crack of half wall l

thickness and the top of SH 2 bundle stresses, GA determined that

this crack would not-propagate instantaneously through the remaining

-tube wall. However, the postulated condition has a crack growth rate of 60,000 hr/in having the potential to cause a leak after a significant number of hours. But using fatigue crack growth data developed by GA, the likelihood of tube failure is considered negligible. -

GA's ' assumed crack depth was larger than the .006" scratch allowance established for the tubes used in construction of the FSV st'eam generators. Therefore, the assumed crack would have had to occur after the bundle was fabricated.

LOSS OF TUBE WEDGES AND SLEEVES The previously discussed conditions considered primarily as-

] designed conditions within the SH 2 tube bundle. Since the SH 2 tube bundle design contains 17,641 wedge / sleeve pairs, it is conceivable that randomly located assemblies have become loose. This possibility was substantiated by several persons who witnessed fabrication and assembly of the tube bundles. They recollect that some of the sleeve / wedge pairs did loosen due to unknown reasons after installation fabrication of the tube bundle. The loose assemblies were reset prior to completion of the module. Loss of a sleeve / wedge pair increases the diametrical clearance between the tube and tube support plate to about 0.125". This could change the flow induced

-vibration effects on SH 2 tubes.

In an attempt to establish a cause for the tube failure, GA has postulated that the larger amplitude for flow induced vibration permitted by the loss of a sleeve / wedge pair could result in fretting

.o wear between the unprotected. tube and the tube support plate. This creates a mechanism for creep and-fatigue crack growth.

-If variations in assembly fits allowed ' loosening of some i7 ' ' "' sleeve / wedge f assemblies which saused the two tube Teilures due tu flow induced' vibration, the failures are considered random since loss of-a sleeve / wedge pair is considered a very low probability event.

CONCLUSIONS 1.,

The location- of .the leak at elevation 4806'-7.6" (4.1"

~

above the: bottom of SH 2) is not near a 3D tube bend or tube weld joint in module B-2-3. This is similar to the leak found.in module B-1-1. Therefore, 3D bends or weld joints'are not a cause of the leak.

j 2. Tube 37 passes through tube support plate FA within 1/16" of the elevation established for the leak in module B-2-3.

Tube 27 in module B-1-1 also passes through FA within 1/16" of the elevation' established for the leak in that module.

I This coincidence exists, but there is no identifiable cause for the occurrence of the tube failures at this location.

3. The fixation of the tubes to the tube support plates is not a cause of tube failure.

4. It is improbable that the leak was caused by vibrations due to transverse flow of helium over the helical tube bundle ,

l of SH 2. Vibrations can be eliminated as the cause of the '

l leak except perhaps for the case where random sleeve / wedge l pairs may become 1cose.

i

..y ,.-%,.___c-- ,,.u-. --.m_-,m,~,.,m_,_..,,.. .._w_., . , , __w7,-. ..m__.,,.,,c,,--%,-..e,%.,-.m-.3....w ,.wr ..e.- ,, .,,, .,v.~,ww-.

' -a ,

5. The effect of carburization of the tube material is not a cause for the tube failure in the SH 2 tube bundle.

6. It is improbable that the leak was caused by thinning of

- ~ - the walls due to wear of tubes 27 and 37 where they pass through the holes in floating tube support plate FA.

7. There is no evidence that feedwater chemistry could be the cause of the leak.

8. Corrosion on tiie helium side of exposed SH 2 tubes is considered remote. The effect of crevice corrosion or corrosion buildup on Alloy 800 by the FSV moist helium environment as a contributing cause to the tube failures at or near a tuce support plate is considered negligible.

9. If the SH 2 tube bundle has been constructed as designed, it is unlikely that low cycle fatigure or crack propagation were the cause of the tube failure in module B-2-3 or B-1-1 unless sleeve / wedge pairs are loose or missing.

10. Even though many possible causes for the leak occurring where it was found, have been evaluated, a specific cause has not been established. However, the coincidence of the two leaks occurring at or near t.here the tubes pass through the same tube support plate remains intriguing. Until additional leaks occur in similar locations, the cause of

'the two tube failures must be concluded to be unknown. _

11. The two capped subheaders, one in module B-1-1 and one in module B-2-3, have not affected the performance of the Fort St. Vrain reactor.

. . - . ._ . -- _. .. - _ . _ ._. _=_ ..

ACKNOW1.EDGEMENTS The authors thank the staff of GA Technologies Inc., Fort St.

~

Vrain Project, for their. independent reviews and coments that helped the authors in preparing this report.

4 e

b I

l f

... p , u p g . .. --..--.---.~~~.~~ ,

"<7 9. % A b d.r.E,AGOR.ARRANGD. R E.,3 ,,

. ..,y.... .

> a.. , ~ _ . f. w - - ,

w. ,g.:p., .. .. e , . < w.cr.segg e.. ,:.; et. .. .. .. .

- + , -

y. w,m::%w

-O .e.w:+.--

MW;c wn . .n .

(;

hP ~ J cana Nkk.h,3.$%i.M 47 . , . ..,

(h i

[...

} MB HNB

,.., x

.M* .. m. ..w.- ..m -

. um H

y ,

w . . ;. .

.%y+ . + .. ..% . -esmut -

orc:.6 r-

$4 MIB W M I '! ! '

@M-v #.. g n N,jl, , .

i l

l

\

p..c,g;ue.G.

. e.. ... ; -mo .

~

nr ypsgra :,

9.jRM.h,p,.

'-. 7 . {.I ,

/ .EEGAMS m

,fb.'s Lt.?f'W'~

' ?c

%am l . Q =~ QiaQ

. .F - ._

w l

- Fest 9 F: munu ifgi,# canum

.g . , .

. .. g -s t

~ ,- - G

i

s. -

1 h, _-

a .

~

A.--.:... f.:. . ~.

i F I G U R.6 1

\

l I

1 i

l 1

l l

i 1

l

o FUEL 60 ACING POETRATICNS (37)

N :=

I d =.- = Q. ac:

; cRivE ANo

, ORIFICE Cou INLET PLINUM

\ s i ' A33EMSUE3 (37)

j

• ~ PCRV b 'n HEUUM R.CW .

h [ h [l i ANNULUS CQlE ec,, <: y pg / rW '4E.!UM g

A I {

OLII

/ ( l CORE q

SU990RT m FLOOR UNER- STEAM

- GENERATCR NM GilCULATOR -

] ,

MCOULE (121 (4) ll NEuuM - -

-m -=== -

-- = u e uM CUT.ET ,

QRCULATOR f( ll l 'A -

SUCTION I m =r' 'O E0 TION PUETRATION--

/ l I f,ll

• V i==:i i

'l ,!

/ -

! - - ~ ~

lI- WAIN sinu lRExursTau HIGH PRES 3URE HOT ROUT TUR8INE DMAUST $;ggy STUM Schemark of the F3V prunary coolant system.

~

FIGorRE 2

I.1/4 94. 0.0.q 47 18. B .S.

UPftt Stun , i

.i -

i, r4

-tiatiiss ais r*t,, p _

.. . , m' M,,, ioos ,

too assa etw arta ---

.. .:, T ll I &

e,s.,s n t.o.

0.

wJ ijf -.b. ./...=, u. niin

.i ..

. i,i.e. C, . . 0. . ne- .=L u.m **

l M.. ' w

sept.wAfta 2 - [ .@@ ,g,3 998 FT I 8 N g '

$h itse S L.60 71 teos ;l 1010**

St f!>0 Plsa ea80e.,a .C .0.GR20$= I Bwall -Q. T litt*F

' I e SUPEIDetAY R I--

2440 FT I 'Bh II "

54 fue S t=171 #1 teos I op. I'"-

l' O. B. te i. M 3 8/4 Ce .0.11$." I so TO 0.305* wall k.

Jr t=

c -

It! Cr . til se.

-; ii r t

, 4..

(CSIIWEIIER q ] ,

}pri.

b. .

p $sec.ue IHM C004 lh LMA SEAL 8' '

.TO_P 08 Lowety_-

m" .

m m e= $7t AfE Le AS WT

, 14 TUtts II + %F Egwaf t.

1 M5 646 8114 l, 8.ema89 CL0suat l

\'. ~

.f i g pg , e fg ,,,

• TO ks.P it cr *cas l l

si n.

f .- s. 'is . ..

,t.o. . ., ,0. , , .t .

i . . . . . . u I a 1 1 1 4

l i,,.....

t a. cant =n=

...,, 00 n,u - ,

at

, .t..t v.. . . .. . .t..

...C...m..

SECO.048, CL0su.t, paafp W fu n

Ct.tasno. $&Etvil $ I,

; ;, .n ,t c.,. . Cro.  ! i an- a ac== ,

ur...

\ ,  !,..r -----*s lj'nu:w. , Cuss aCe l:

Rii""' """'I

\' W./ tig d_.J,_

w  ;;,$"-

no .. gy  :

r. n.: i.. i F, t.t0.,4TI.

E S. ..

.. \g__,

I'I.' . .'I..  ; ;

5 attDwatt. pe pt .

. (W Chan 6 l pawta repe #N

• k *"

(tLLOwl CC..ECfic.

0 -C 0 w'"14 '. '

. . .. . . .I. : .

3 C0tg 9:Lw:wl .t.ta.f.(Ci t 0.- 1 COLO .t.taf ..st?

Can . CD

.e

,P--u- -* - 67 s*r i

(0 =ta, Di et

] h1 P%it

,.t....._ q 3 i g ex.0..t

n. ..t..f... n ou ..t u. ,E w.. . . .

. ornut.,=t o . e. oiCaet0 wist

~ . Cm.se. ...

EC a. 600 *s a icot'r S.u' a. .<f. CL. ass.tG.*.M t. 't . - _ _ _ . , . . n

. . C ,o E .. ..w0 C -.t . . . . , ..t.,t, TYPICAL S* REAM Gggqq g FIGuas r 3

- a .

L;

.?

b~ 9 j '

I G I-I S' .~p s w,ir '.e  :

!l D!

9;. .

d ,

~4 M

.D 7.w  ;

l r

Tl r r

i r l

~.r:.2 .'-

l

jk; i

Fig. 3 A steam cenerator steam section Being Assembled to Primary Closure l

e

....a...+ . I w w. .mpyx y en c -

1 r.g i.hk # b ,eIe.N b I 9.$5EbIN !k -

3hik,'.'h.Nhl'1/f././;'.,f[ d. h kih sq-=g;;,.'s. .

, ~ d 'g j!,j %jg g

. >,li A " . 2y i '. r . . .. ' . &,i 5s1 llj i,,l{, gl %g ,i ihM

- . . 3

.I . 4. y. .

.d-At =~. ,- e sv, , . 9 c. .. 3 .

y 4 ,- -

- ['f$,.. ,, ),'.

ylii up;p3g [ gg 'N 3

9.J, - [ at j Q 's g 5 ' ,% ""A.i.t.:..:}.y. 0((* jh,h

,y c i., O gg C'm '--= ~~' ,- l - N M, pg,65,*T* .i, ii g sw

v. s 4'

F3<e -

}i t

g

=

g j jj ji. e

'ii.

m i.ei M

-r--'a'*'

stattrhw it _

L'ag/3, da4 eM*

u u*. .

,. 1 . n.e, . .

- - .p -

I,!s,.,,g~s

,]! l s - . 2, . m, w+ -

. .i,3 i.g.r.s 0 ,, ...

1 s

ip .

a.... <'

g y ,. , '.

3 ., i 3 3 ) o , =. - .p, .' o3L  ; .....;;;; il !

~,

d

. . -. .,3 11 -

tdi is, ?n-le .

5 1

I! . ' . .-. . ,s. . .

- . g y n

~J <

] lk / / T r i ; p ' lllN h [i  : .."I L .a n.,N !

h ; 4p ti e - o .

. i , ,s ,.1 5

.., s it .,

e i GPn,1- :- ::::: - ' == h<-1s. 5 -

d i 8,[!  ? E*

p1;.j,ds1i[13 j }O -

I-- jit .  ;

. 'I*. - *.*.*.

,n. . . l3l x 1- _

l{1.

y 13 h I-g pj a ;uiInM 1

jl s l l Llj [i 3 fA NJe NIPI d ~ 'd 3.l2 -

i

.j =@ TJ i

1, ;]

. s.

wsAj . .,, w. 3 .-p1 t

, . e

,-. 3; A~e .n..

.- a n

g e

.i

, g s.g;13p. . . . ,,5- m,: -

j i ..-

. +,.. s.e.e.n.

a 1 w a

, , t i' . , , _.z_ue

,r jI .i - 4 _. _ .] ,  ;- \ - n a -.-

t,

, y _

==, W..P' T, U l g , ,i t I g

t 'h*

a;j . -! a ;!sIjl 4_m n i

h.

d il $.~~ p T. 's r

9:

ll "a** J 1

i! !

Ll j

Ja 1 l i. '

j *:..:A **

53 a Mr 1,1v.((g ,li 1.11.!, II:. td ..

.- - is  :

y

=un= *h a olilb-.

a y Id" d

.a, i.,

I

,k.,i$

1 x) it .

a 3 I

, . u ne

• • 4 i 10 m a. M ih I' l - ~- gx l7 m ~/ 2]f i r ' \ *!*!@ s ft ;g t ~

3 1 '. . se - e a:

hg@.Skg$r,./ /

i >

/ 'E 43A

! ,g --

'$ I L ipt  % . N .=, s.J.>. M... 1.me L.. J ,t, ,(z n. .,L.

i 1..

t . ..

u '

g,, t/ p/ .ices......o ,,,..> ui ma +:g ,j , --un. u, : \v:::n' my n

. . n.

n.M r r 4,onc j[ 9,

,, .y

a. .lt hp g 3 3. -h-? .h3Pg. .

F, r, i 1*  :

.c. ' \n*i.!.- I I. c ,

. .~ .J.

u [-

q3 1 -as 3 *;

sa iggg los  :

n< . l,i\.

s ,.a -u .

a

. (u

. I ,, _-e eset -

e-snt, ,

s

• j j

1 Wa.

g.

l. ' ll l

9 ;1,  ; s 11 ..

(;

~

h a,! 3:

.. $ j, .'i ' 1, '.'

r# .

d ) li i .L . y , o 1

.g . . . ... w, >A ,

g y - --,,,_--,,4 , p,-- .,-,-g- - , - , - -

-en.,.w,.., - - , . , -

9 9

e e

S Y

-A* M:

i!

m _

se:oncey m

.m. - 4

- . ~

n e ux s.rrve jg 3 i" AM SE: NCANY STEAu O ER. ton inTunspect - l g y ,7, w w w

~ ~.  : ~

k SU4MEACEN ?get Tysg m cap - t hf STus

• = 1 ru.c =

. .=,_,

pbMN~. -- .

7YPICAL SOSHEADEft. TOBE, c,AP F IGoR.s 5

F1LL. FUMMEL % \ /

A sit'-0"ELEY. - _

7 o 'pr TOP OF 5 4 2 ,

. 1657" TAFE

/

$H7 MEASUR.E i

TuSE SUNDLE ISoLAT1ow VALVE SER.vicE .

1. AR.GoM gas EES ,

2. HELIUM GAS TUSE euNDtE N .

LEAR HOLE

5. GAS LEG AT 51 4 "

4 GAS SOPPLY VENT

[ .

G, TEST RIG .

6. WATER. LEG

1. EES LEG .

8 SHt. LEG .

9 GES LEQ DRAI A "

' 'k"0D f a SH 7 LEG DRAIN .

-.15FT Hauum w '

\\%\\\ \

F-( ASK

$ >N 'FRJNARY CLc5dts 2 nY # - A i 3 1 (Y %K\\\\\\\\^'

1>d-- L x'\'d f6 o 0; '5EcodDARYCLOSCRE:

X5 7 f f 8 I><I X ARGON pmg 4h [9 '/4"oD cu.

\ k

/[co Co. [ 10 p , 'TusiNo rus:No l

FEECIMATER. SOBWEADEK 13 MAtw STEAM SeeHEADER. N-96 l (CAPPED) (CAPPED,) -

1 l

l Test RtG Fca FINOMQ ELEVATION OF TO!SE LEAK IN STEAM Q EN ERAWR M cDutE S S l

FI Gog.e G

- _ _ . _ . . _ . . . . . . _ _ _ , _ _ . _ _ _ . _ _ _ _ . _ _ _ _ _ _ . . _ _ _ . _ _ _ _ . . _ _ _ _ . . _ _ _ ~ . _ _ _ , . _ _ _ . _ _ . _ _ _ , _ _ _ _ _ _ _ _ _ . . . _ _ _

h F 1 G o ra.e ~7

'T 6st DATA coRue Osm4 (n ma suosersa. ( Roue.s Q Lt rstes or- toATBr4 DRAIMEC F(f CM LOATG1 L6C3 70 7f SO $f O So

• E 8J lb _ __.

o K, socrem esse 2 coic (4.aoe'-3"e tev mov)

' 57 "~

. see aw.e e nec coic sum Arrues w a

b Go '

J Lil hJ i g G 4..s " .

2 \

las tl W

4 y 'lo l 3 i

w d <

e W

.T U

80 90 '

A 9

. ,~w~~ . . , . _ . , . , , . . 3.

i.13 g

' 7

~

.I ._. h:dh 1- ..

s 1 i e # 1- , 1

d

.hg

!i !f

! jiMM-M4#s ,!!'![!

=

!YTWi-!#HHHe

~

l ,! r i l .!

! i!II f1l :hl i!

b Ull ig.. .i., . ,

r!;l,'

i., ((

.j j n w> ..rim m

ri j e (,

[.g liIif l "

q,ji $' Idi$

- ij .. 'r.. Ii jtj 4 :. -t iR j.!

ia $;T@'mr ai fur w

e -

> !y I ,Itr r[e;y1 i:v i iW!>!"' La !aiiii 3

• '! 1 @NI$5 LI f,

~

e(, G. ,p1 s

s. i.

3gj '

Eg a* .

.8N76hilE...

- i .s -

y

,ig tp ac

!ikl -.

,w' -

S-e . -,.s.. < , . .

k

- \ '

/)l l

.  ?

m .1 i $

. g ,

,j s= === e . - 3 f' 1x ' J,- -k, l. n)

' 9 ==~.= L a 6 r

) ./ 3

/DW} H

\s5 o

J .

!! :st

~

D w

" "l -Q

• u 9 ,

0 t t

8 e o t

! i a i

t i

! ' ,i,*

! F l

'V4 $

\

^

NG '

l f f-f,l ,

i e r

i

END OF COIL ELEVATIO N 4 B O ta ' 3" W HER E WATER LEVEL RISE 3 DECREASES SIGNIFIC ANTLY / h i FOR EQUAL INCREMENTS 4- TANGENT r OF W AT ER. ADDITION. 4

/* J

• * " '- S E*~

( ggggxxxxx\\\\\\\\\\\\\\

1 s

/ gggg\\\\\\\\\\\\f\\\\\/ -

d . 2 0 5 WALL 1 . G .3 0" 1. D. TUBE I.000" Q.D. TUEE d

A 9

a

! l J

e END 4

g BEND AT THE O I: THE l B OT TO M COIL OF S H . 2.

HO R ZO N TR d r _-

( C Y LI N DE F" II, TUBE 27 / 37) 5 \

_/

j ,

REFERENCE S-R DWG N O.

21610 -8

d

\ '+ 1 -

D O W AIC ord E R FULL SCALE

\ i FfGURs c3

TABLE I NOMINAL PLANT OPERATING CONDITIONS Full Load One Quarter Load 100% Steam Flow 25% Steam Flow

• Load Generator Output (N) 342.0 81 .2 Station Output (N) 330.2 67.4 Helium Flow (lb/hr) 3,410,040 973,440 Outlet Pressure (psia) 686 589 Inlet Temperature ( F) 1,427 1,236 Outlet Temperature ( F) 742 569 Pressure Drop (psi) 3.60 0.37 i

Feedwater/ Main Steam Flow (lb/hr) 2,305,320 576,360 Outlet Pressure (psia) 2,51 2 2,419 Inlet Temperature ( F) 403 299 Outlet Temperature ( F) 1,005 1,000 Pressure Drop (psi) 590 Reheat Steam Flow (lb/hr) 2,245,800 556,920 Outlet Pressure (psia) 600 151 Inlet Temperature ( F) 673 570 Outlet Temperature ( F) 1,002 1,000 Pressure Drop (psi) 42 11 -

• NOTE: At low power, required helium flow rates in actual operation have been found to be greater than originally anticipated, re-l sulting in over blowing the core. Under these conditions Technical Specification limits on power-to-flow ratio are al-ways met.

3 t

i STEAM GENERATOR MAIN STEAM TUBE PLUGGING SUP94ARY

Additional Feedwater Ap Required to Generate Full Power

• Hot Helium Excess

! No. of Plugged Subheader Temperature Over Plant Subheaders Madrle Irisi Trim Valva 1427 F Design Load j Case Per Module Valve (psi) (psi) ( F) (%)

1 1

1. Al1 nuJules 1 40 25 0 100 innermost coils 2 150 90 17 100 pluJged 3 250 119 34 92**

i l 2. Single module 3 125 115 6 100 innermost coils plugged

, 3. Single module 1 40 0 0 100 plugged coils 0 6 100 2 125 dispersed 3 240 0 12 100 i

i Available excess ap is about 250 psi.

i

! Conditions at reduced load correspond to the nest severe plugging combination possible.

I l

l TABLE 2

L T. A S L E 3 STEAM GENERATOR MATERIALS

. Pressure Parts Section l Material Designation Main steam / water Carbon Steel SA-210-Al tubing and piping C-1/2 Mo SA-182-Fl

'and associated components 2-1/4 Cr-1 Mo SA-182-F22 2-1/4 Cr-1 Mo SA-213-T22 1/2 Cr-1/2 Mo SA-213-T2 Ni-Fe-Cr SB-163 Gr.1 & 2 Ni-Fe-Cr 58-407 Gr.1 Reheater steam 1-1/4 Cr - 1/2 Mo SA-335-P11 tubing and piping 2-1/4 Cr - 1 Mo SA-335-P22 and associated components 2-1/4 Cr - 1 Mo SA-182-F22 Ni-Fe-Cr SB-407 Gr.1 Ni-Fe-Cr 58-163 Gr.2

~

. Major Support Main steam / water 2-1/4 Cr - 1 Mo SA 3870 Structures sections Ni-Je-Cr Ni-Fe-Cr SB408 Gr.1 & 2 SB 409 Gr.2 )

Reheat sections Ni-Fe-C r SB-409 Gr.2 C- Si SA-442-60 Primary closure 2-1/4 Cr - 1 Mo SA-182-F22 Ni-Fe-Cr SB-408 Gr.1 1 Cr - 0.2 Mo SA-193-B7 & SA-194-B7 Secondary closure 2-1/4 Cr - 1 Mo SA-182-F22 Ni-Fe-Cr 58-408 Gr.1 C-1/2 Mo SA-182-Fi C-1/2 Mo SA-204B Shroud 2-1/2 - 1 Mo SA-182-F22 Ni-Fe-Cr SB-409 Gr.1 l' Cr - 1/2 Mo SA-387B Baffles and 2-1/4 Cr - 1 Mo SA-387D cover sheets 1 Cr - 0.2 Mo 4130 Ni-Fe-Cr SB-409 Gr. 2 _

Insulation Kaowcol Kaowool Silica Felt Silica Felt

-)

. ~ . _ _ _ . . _ , _ _ .

1

) .

A. LM*4Ce ITf2tfgt u

, Tues HELIX Lik -

" *"N

.MO. PlTC H PLATE A PL ATE FA PLATE 8 PLATG Fa PLATE c. 'PL ATE F C tsu6ru WedEfWL of- END j c.<

• E lev. a<" ELEV. e<' ELuv. ot* ELEV. od' E.LE. U . ot" ELEV. Coa t

LP O

{ 37 8.63* 53.36" l08 "Z.. S 9 " D3 4.I4" 118 5.4 6 tag 3 *P,02 " 3 43  %.3 4" 4l3 9,9 o

• So,/a LP

-38 6.98' 4-8 .34 102 4 '

167 , 3.'t s " 12.2 4,26' 2t'1 6.SI" 342 6. 56 4er ~7,28" 6 0.*4 "

LP

!. 39 62S* 31 4-Y 9'  % Ibl 4' 286 3.82

• 298 3M 336 4.93 48i

• 68' '2 "

L. P f':. ,

1E *).le " S6.30" 19% 5.ol " T6'5 6.45" 382 Bi~o4 " 23 M- 78

• 14 1 3.41" 70.3"

$ LP i is 4 a 8 21.53'- t 92. 2.ss" 25/  % 31 4.I4" 17 -y- ~rt E.74 " 137 + 4 i .S

O LP 17 S.63 S 3.86" 196 4,46" 26: 4.o2* 300 7,33 4 li y 66 .y 131 3,g4." 60. <t "

LP 41 9. 6 8 53.eu* 84 -V I4M 3 .7)

• to4- 269 6.tn* 324 9. f i 73.t

5.tr 29 1 LP ' '

4-1 4ag ** 29.53" 73 4 ( 4-3 4 :37 2.6s 8< 263 ,w 39 4rr1.* 23  %' sg.oz 8.

I l c,( * = A rt.C Frzom DotoNto.wEtt To TOBL. soppe.tr purs More s i REFEenincG S'-(C. Pk3 C,5 71610- 4.,5 d %

Etev. c Tusts 37,39 k"59e s e e40 pott 15 2-3 inews Anavsi I5orres os $1s t (4 Woc.'-3( *)

4

= Tuin rues Nur PASS TH4d THIS Pa.At E AT Tills M ' 70hg 21',16 b 7 1 iM MoDetE I3- l ' I O " E t e: v. ctoses r To FicarcaEta LEAK blEV. Tuar.c 4 i { 4 z, in mons.s IS-l -6 LP = Lns s- PuirE la cast 36 Forte TortNeur3 nota H T ABL E 4

. J . .'

I'ArFAc4Mssr 1 rc7-34&L8

- ( a c a. w es- 5. 5 . w p - s m Q Public service company ce cchesrso

-O . 12015 East 46th Avenue, Suite 440; Denver, CO 80239 i

r March 31,1980 Fort St. Vrain Un f * *!n . 1 680064) fir. James Miller Chief, Special Projects Branch '

Division of Project Management U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Docket No. 50-257

Subject:

Fort St. Vrain Inservice Inspection & Testing

Reference:

1)PSC Letter, J.K. Fuller to S.A. Varga, P-79289 Dated November 30, 1979 O 2)NRC Surmary of Meeting Held August 20, 1979, G-79189 Dated October 20, 1979 Gentlemen:

Enclosed for your review are the Helium Circulator and Secondary Coolant System submittals, which make up the remainder of the Category I inservice inspection and testing program submittals for the Fort St. Vrain Hign Temperature Gas-Cooled Reactor. These submittals contain draft modifications j

to the FSV Technical Specification Surveillance Requirements, along with an evaluation of the proposed requirements, as in the previous l submittals.

Enclosures (1) and (3) contain the draft revisions of the FSV Technical Specification Surveillance Requirements for the Helium Circulators and Secondary Coolant System, respectively, which include the following sections:

P G 0 n 4/ W'/3

c' l

l . . - . _ . _ . . -.. .. .. .. . -- ... .- . .

l

• . t

.- 's Mr. James Miller March 31, 1950 a; p=9e 2 -

a Enclosure (1) m _. SR 5.2.17 - Helium Circulator Pelton Wheels, Surveillance SR 5.2.18 - Helium Circulators, Surveillance

, SR 5.2.19 - IACM Diesel-Driven Pumps, Surveillance SR 5.2.27 - Helium Shutoff Valves, Surveillance Enclosure (3)

SR 5.3.1 - Steam / Water Dump System, Surveillance SR 5.3.2 - Main and Hot Reheat Steam Stop Check Valves, Surveillance SR 5.3.3 - Bypass and Pressure Relief Valves, Surveillance SR 5.3.4 - Safe Shutdown Cooling Valves, Surveillance SR 5.3.9 - Safety Valves, Surveillance SR 5.3.10- Secondary Coolant System Instrumentation, Surveillance The revised portions are identified by vertical lines on the right-hand-side of the pages.

Enclosures ('2) and (4) contain an evaluation of the additional or modified surveillance requirements which are the result of PSC's review of the current surveillance requirements for the Fort St. Vrain Helium O Circulators and Secondary Coolant System. The systematic approach used to develop these requirements is explained further in the first pages of the rev.few.

These four enclosures complete PSC's review of the surveillance requirements

. for Category I systems and equipment. PSC is proceeding to review the inservice inspection and testing surveillance requirements for the Category II systems and equipment identified in Reference 1. Prior to establishing a schedule for completion of the Category II review, PSC requests that the NRC complete its review of the Category I submittals and provide PSC with comments on the adequacy and acceptability of the approach being utilized by PSC to establish detailed inservice inspection and testing program requirements.

Also attached (Enclosure (5) is the information requested by the NRC at a meeting held on August 20,1979 (see Reference 2) concerning Fort St.

Vrain equipment built to ASME codes, and concerning inspections and tests being performed tilat are not currently included in the Fort St.

Vrain Technical Specifications.

4 6

... .m -

-m.. w . - - . . . . -- . ..

4 ---- -r - - , -

• ~ -

( ,

j. Mr. James Miller

{i A V

March 31,1980 Page.3 l

4-i Please direct any questions you may have concerning these submittals to Mr. Mark A. Joseph, (303) 571-6671.

Very truly yours, O

J. K. Fuller, Vice President Engineering and Planning JKF/MAJ:pa Enclosures (5)

I O -

._, ... .. - - +==+= -

*ew--

-re-- --

u+--~ .- erwe --e w- m-4---,-

r ++_-w-= m---u+-* - --A er- v- --

.. .- ,/ -

. ' c:. :. .

1 q

.. Enclosure to P 80064 (4) o LO i

FORT ST. VRAIN INSERVICE ~ INSPECTION AND TESTING PROGRAM J

J

! O I SURVEILLANCE REQUIREMENTS REVIEN FOR THE SECbNDARYCOOLANTSYSTEM (22)

O- pcaoo!'Y EE-22-0002 Rev. 1 March 25, 1990

' ' ~ ' ~ - ------ _ ._

N g 9 e-tma F$yveY} N

$ $h *F T T'-- 1 f'*7 =- - - - * -

• --'r-- w-

• -r a .

( -(

,. v EE-22-0002 Rev. 1 q . p.19

4. . STRUCTURAL INTEGRITY t

j 4.1 -STEAM GENERATOR TUDES q.

i a) - Current surveillance requirements:

p-

, There are no current technical specification surveillance

.!! -requirements for the steam generator tubes. However, existing plant instrumentation continuously monitors the steam genera-tor-heat transfer sections for evidence of tube degradation.

Moisture. monitors installed in the reactor instrumentation and analytical instrumentation system will indicate, alarm and initiate necessary protective actions if a leak should develop in the superheater section. Radiation monitors installed in

- the secondary coolant system hot reheat piping will' indicate, alarm.and initiate protective actions if a leak should develop in the reheater section. In addition, instrumentation in 4

the analytical instrumentation system and the steam and water sample system continuously monitors primary and secondary coolant purity.

Existing plant instrumentation also continuously monitors the superheater sub-headers which are located within the PCRV penetration. Penetration pressure and moisture monitors i""** ** *" *** "c^v ^"*i i^rv "v'**= i"*i""** * ^== ""*

' O- initiate protective actions if a leak should develop in one of these sub-header sections. Reheater steam piping headers located within the PCRV penetration are continuously monitored as well. Penetratiens purge flow monitors in the PCRV auxiliary system alarm and initiate protective actions if a leak should develop in these reheat piping sections, which are at a lower pressure than the penetration interspace.

b) Recommended surveillance requirements:

i The once through type steam generators at Fort St. Vrain operate with reactor coolant (high purity helium) outside the tubes and secondary coolan: (high purity steam / water) inside L the tubes. Thick wall tubes, arranged in helical tube bundles, provide the required strength and flexibility to withstand the high operating pressures and temperatures.

i Structural integrity of the steam generator tubes is required for safe shutdown cooling, where either one sucerhea:er

- or one reheater.section and one of the corresponding circulators must remain operable or be restored to operation within the i time delay allowed before a permanent loss of forced circula-tion accident is declared. Structural integrity of the cubes is-also required since they comprise a portion of the reactor coolant pressure boundary and therefore function as primary

()

containment, one of the major barriers against relcase of ogee ,eiig: _p ie _ e io- e ea ~ ---we o.

~

4 .

(m (.

.

• EE-22-0002 Rev. 1 p.20'

, 4.1 (cont.)

fission products to the environment. - The tubes also act as 4

. a barrier to prevent steam or water from entering the reactor coolant system and possibly damaging reactor internal components.

'There currently is no method available for inspecting n- steam generator tubes without removing steam generator modules

from the PCRV. The tubes-are not accessible from the primary side due to the shroud design which surrounds the tubes and cannot be inspected internally, using current technology, because of the tgbe. design (helical tube bundles, varying tube I.D. and 90 turns at the tube to header or subheader junctions). Although the PCRV was- designed with provisions for removal and replacement of steam generator modules, it

' would be a difficult, costly and time consuming task. Further-4 more, the method has not been demonstrated nor is equipment ,

available to do the job. The'refore, non-destructive examina-tion of the steam generator tubes is considered impractical at this time and cannot be used to verify tube integrity.

There are no provisions in the design of the PCRV for insercion of tube material specimens in the reactor which

, would provide meaningful and representative data regarding tube integrity since the primary concern is cyclic thermal

'y stresses across the tube wall. Steam generator tube integrity must therefore be monitored indirectly.

O Corrosion protection for the steam generators is dis-cussed in section 4.2.4.3.8 of the FSAR. Therein, it is shown that-expected metal loss for the tube materials operating at specified conditions is insignificant.

Corrosion in boiler tubes, in general, is related to feedwater quality. Out-of-specification feedwater chemistry

, could lead to excessive deposition of contamination on the

! inside tube wall, which could accelerate corrosion. The l results of a special program to monitor steam generator per-formance with respect to corrosion were reported in reference

18. Data monitored included tube side cressure drop, heat transfer section temperatures and feedwater chemistry. Data analysis showed that steam generator performance was as ex-pected; no evidence of deposit buildup was observed. Reference 18 also reported the results of examinations performed on two l

feedwater ring header trim valves, one feedwater ring header stra.iner and portions of the feedwater and steam leads which were removed to plug a leaking tube. Again, no unusual corrosion

, conditions were observed. These findings tend to verify the i design expectations that corrosion and subsequent tube thinning is of little concern for the Fort St. Vrain steam generators, i provided that primary and secondary coolant purity is main-tained within specified limits. Since the primary and secondary

' es coolant chemistry is continuously monitored there is no need

! (s/ for additional corrosion surveillance and none is recommended.

,,, e . - m- m em . .* * - = = -

eer- M q r yga y e-. - ---

_.y---p, ..y. y , ny. --w-- g y-,-,p ,m ,,oy-,,.y p.- .. yy 9 g- m -

W9+ wis 4 -

_r y , .

y .

EE-22-0002 Rev. 1 p.21

~.

1 4 .' 1 (cont.)

O i

- Vibration and wear protection of-the steam generator i tubing is described in sections 4.2.4.3.4 and 4.2.4.3.5 of i the FSAR.. The design provisions for wear protection are based on' proven technology and do not present any concerns with respect to tube thinning. .The design provisions for V.~' -" -

vicration have been confirmed by the plant startup test pro-p gram wherein one steam generator module was fully instrumented with strain gages.fThese measurements have not indicated any cause for concern relative to tube vibration. Therefore no surveillance is considered necessary to.further monitor for tube motion.

Design of the tubes to withstand the stresses imposed by operating conditions is also described in section 4.2.4 of the FSAR as well as in the steam generator design report.

The-analysis met all applicable Code requirements ia existance at that time. Since then, several potential concerns related to tube life have been identified, based on more recent knowledge of creep-fatigue interaction and cold working of Incoloy 800 tubing. Additional research and development pro-grams are necessary to determine if these concerns could 1 actually affect predicted tube life. However, since the tubes 1_ cannot be directly-inspected, defects can only be identified after they have prepogated to the. point where a leak is initiated.

'n A review of the current design provisions, as described in (a)

v. above for monitoring tube leakage, indicates that there is sufficient instrumentation provided to detect both small and '

large leaks. . Technical Specifications LCO 4.2.10/4.2.11 also specify limits for primary coolant impurities, which would include moisture entering the reactor via a leaking superheater tube. Technical Specification LCO 4.3.8 specifies limits for secondary coolant activity, which would likely be indicative of a reheater tube leak. These provisions are considered adequate to verify the integrity of the steam generator tubes and meet criteria 3.3.2 (c) of reference 1 for continuous leakage monitoring. Surveillance requirements for the instru-mentation and control circuits which are used to verify tube integrity will be addressed in separate reports covering the plant protective and radiation monitoring systems (System 93).

There has been one superheater tube failure to date. It was identified by the instrumentation described above and was subsequently located and successfully plugged. The vertical location of the leak, as determined by indirect means, was not located where failures would be expected due to the potential concerns previously mentioned and was therefore considered a random failure. Experience to date therefore does net justify that additional surveillance be considered.

- Regulatory Guide 1.83 has been reviewed as a reference (j

_ document, to determine if some requirements could apply to the

. Fort St. Vrain steam generators, even though the Guide specifically addresses inservice inspection of PWR steam gone.rator tubes. It was determined that the differences in design are such that Regulatory Guide 1.83 cannot be considered

. a nni i <-a hle. . . _ .- ._ -. _ ._ ___

,_  ; \' , ,

r-

' t..

a v EE-22-0002 Rev. 1 p.22

- 4.1 (cont.)

c) Proposed ASME Code requirements:

The inspection and test requirements of subsection IGB would apply to.the steam generator, which is considered to be Code Class 1 since it forms part of the reactor coolant pres- -

44 ~ sure.buundary and functivas as. primary containment.

. Paragraph IGB-2510 requires components to be examined for leakage following each reactor refueling outage and to be pressure tested and examined for leakage at or near the end of each inspection interval (approximately every ten years). The required test pressure for the periodic leakage-examination is nominal system operating pressure at 100%

rated reactor power. The required test pressure for the system pressure test is 1.25 times the system design pressure.

Paragraph-IGB-2520 and IGB-2600 require that components

'be non-destructively examined for defects as specified in Tables IGB-2500-1 and IGB-2600-1. There are no requirements specified in these tables for steam generator tubes.

Subsection IGX, which is not yet available, will provide requirements for components subject to elevated temperature service, which would be the case for the steam generator tubes. It is understood that these requirements will generally 4

J{}. provide for surveillance using material specimens.

d) The current and recommended surveillance requirements exceed the proposed Code requirements for leakage examination, since leakage is continuously monitored at power. For this same reason and since the tubes are not accessible for leakage examination, pressure testing would not provide additional assurance of integrity and may not be practical for this design.

4.2 STEAM GENERATOR CLOSURES The structural integrity of those components of the steam generators which functions as primary and secondary closures

. for the PCRV penetrations has been reviewed in the reports on the PCRV and the PCRV Auxiliary Systems (references EE-ll-0001 and EE-ll-0002).

4.3 STEAM / WATER DUMP TANK STRUCTURAL INTEGRITT

, a) Current surveillance requirements: None.

l i ,

t

. . - - ~

- , - - , , - - , . . , - - - ,.-.w_,y, . ,,, r,-, --. , , . , -

g.q_.- y--__mg_ ,,_,*"yyrt--Mr --+--M y9 -y PNe "--<Pr--mte r--

E _ . . . -- .

e ATT AC.HMENT -3 Tr P-84 02 3

• (Excs.tvrs Fibm G4 2.o r p) s :ts =tv m-urs. January 5,1982 g h gg = aw ve ro. Q-13:82 :5 g.3y,,'-)

c:cx gsc3 (ge j~Up I.~

Les Alamos Naticr.alLaccratory wro.

tea. oss.

576 (505) 667-5150 MAMEX. W. J.

fiF ER: C,4

.n . .

d-t Los Alames.New Memcc 87545 r HILL. E. *. ~

l set; ts. J. 9.

i JcMNS. OE."

. Energy omsion g-go/9' .

L.II.

, cm:s.a. R.

tu.a. c. x.

L. ,.

Mr. Robert Clark, Chief FW43 -

QJi scic cLtd tEC:R3 CDUA Operating Reactors Branch 3 Division of Licensing @hgh 2,g:t (:

WMEMScuR3. 3. E Office of Nuclear Reactor Regulation W N D. !-

i Nuclear Regulatory Commission lEW ER CCC. c0M.n Washington, D.C. 20553

Dear Mr. Clark:

Enclosed are two copies of the report concerning our review of the Public Service Company of Colcrado proposed Inservice Inspection (ISI) program for

=l the Fort St. Vrain High-Temperature Gas-Cooled Reactor. The work embodied in the report was carried out by Advanced Science and Technology Associates l (A5TA) under subcontract to the Los Alamos National Laboratory.

' We have reviewed this report and concur with the findings and recommen-dations specified by ASTA. You may wish to direct your at antion initially to the Executive Summary that gives recommendations concerning the Public Service O~ Company of Colorado (PSCo) Inservice Inspection program submittals for the five safety areas that were reviewed.

We are also forwarding the report to PSCo for their ccmments. Except for f

minor clarification that you or PSCo may wish to have included in the final report, we consider.the milestone on the ISI task to be met. If you have any i questions, please call me at FTS-843-5150.

I I Sincerely yours, 8

.&M['

Charles A. An- acn, Group Leacer Advanced Engineering Technology

Enclosure:

a/s i J. Jackson /M. Stevenson, 0-CO, M.S. 561, w/o encl.

'I R. Haarman, EP/NP, M.S. 671, w/o encl.

l G. Kuzmycz, NRC, (1) l L. Brey/M. Holmes, PSCo, (3)

$ t()Y

}

.Q 30 2

An Ecuas Cocortunity Emciove</Coerstec ty Untersty of Cas ornie

-- . . , , . , , ---r--- ,.- - , . - - , . - . . _ .

-- _ _ = -. _ - . _ _

l'o i

O

.^

4

'i in REVIEW OF THE

!O PUBLIC SERVICE COMPANY OF COLORADO I

! PROPOSED INSERVICE INSPECTION PROGRAM O

I io i

i I

. ,D Submitted to:

l

^~

l LOS ALAMOS NATIONAL LABORATORY

.L, s by:

ASTA, Inc.

1 22 December 1981 9 Under LANL P.O. 9-L32-9055X-1 s /

rJ

.... Ql3

'IU 3

q)O//

1 8, .

S~

~~

C0FTENTS

- iO PAGE

)

l

{, i INTRODUCTION 41 O TASK OUTLINE 11 PROCEDURE iii O PROPOSED ISI PROGRAM - SECTIONS iv CODEDEVELOPMENT, ISSUANCE,ANDEkFECTIVITY vi 3 TERMINOLOGY vif EXECUTIVE

SUMMARY

E-1 SECTION 1 - Responses to Proposed Technical 1-1 Surveillance Requirements for PCRV Auxiliary System Tables 1-10 D SECTION 2 - Responses to Proposed Technical 2-1 Surveillance Requirements for Prestressed Concrete Reactor 4

Vessel (PCRV) l D Tables 2-12 SECTION 3 - Responses to Proposed Technical 3-1 i Surveillance Requirements for

PCRV Internals Tables 3-9 l

O N ST

=

l e

y) .

1 CONTENTS, Continued ,

!O .

j ' -) PAGE

%. l Responses to Proposed Technical

[j t SECTION 4 -

Surveillance Requirements for 4-1 l

Primary Ccolant System Helium O Circulators

! Tables 4-9 i

I SECTION 5 - Responses to Proposed Technical 5-1 3 Surveillance Requirements for Secon ary Coolant System

. Tables 5-11 9 APPENDIX A - Additional Surveillance Requirements A-1 j REFERENCES A-36 Tables A-37 4*

.l g

i '

l I

t

' p,D m-1, D

~

t1 I

l

r. -

S -

. . - ,.. . . . n . , , 7 ; - . - - - .- u_ i .v. . - : - c .- - - - - - - --

47?. - , . ;. c.

r---

7

. .)

i <

1 1<

O amN0u A j O ADDITIONAL SURVEILLANCE REOUIREMENTS Identified by this Appendix are areas of inservice inspection i

interest considered to warrant greater attention in order to meet the

'3 intent and scope of the proposed Code. These include areas addressed by PSC in rationale supporting the scope of surveillance proposed by the SRs, as well as related areas not identified by proposed SRs.

O Reconnendations for additional surveillance are, in most cases, predicated i on the outcome of prior investigation by PSC to determine applicability; consequently they are not expected to result in automatic adoption.

+3

, Included in these additional areas of inservice inspection interest 2

are the following structures, systems and components:

O 1 A-1 PCRV Penetrations and Closures

. A-Z PCRV Thennal Barriers

~

'O " A-3

. Core Lateral Restraints 2

, A-4 Helium Purification System l -

_ A-5 Steam Generator Tubing i

A-6 High Energy Piping l

l O

o -

A-1 9 - ,m-.,,,n- ---. . - , - ..u- , . ,, n, ,,..s e - , , , , , , - - - , - - . . - , , , , . - , - . . - - . , - , - , ,

0 Q A-5 Steam Generator Tubing IO

a. There is no present program for periodic surveillance inspection of naam generator tubing at FSV. The primary 0 side of the steam generators is continuously monitored during normal operation for water ingress. Presumably, any tube leakage or tubing failure would be detected by the primary 3 side moisture monitoring system.

b. ASME Section XI, Subsection IGH requires surveillance tests 3 of steam generator tubing materials.

NRC Regulatory Guide 1.83, identified in NRC letter to PSC I

iQ dated January 15, 1981, requires periodic eddy current

examination of steam generator tubing in LWR's.

O Reg. Gufde T.121 establishes tube plugging limits for LWR steam generators.

' Standard Review Plan 5.4.2.2 discusses steam generator sur-veillance requirements for new plants.

O c. FSV steam generator tubing is generally inaccessible for tubing inspection both due to lack of physical access to the iO

'3 A-27 e- --wP

m

] 'D O.

i

,) tubing area and unit configuration. The reheat tube bundle at the top of the steam generators is subject to the most

{

j severe operating conditions. Helical tubing in the main g section is subheadered and is connected.to the superheat section at the top through a series of bi-metallic welds.

The superheat section is located inside the helical bundle.

Bimetallic welds in the top of the steam generators between

)

the main and superheat sections are subjected to severe l thermal conditions. One failure has occurred in this g region.

l Having two (2) steam generator loops assures adequate safe shutdown cooling in the event of a failure in one loop (each loop containing six steam generator modules). This

', assumes that Tocation of a steam generator failure is correctly diagnosed and isolated.

3 Operation of the Steam / Water Dump System and Moisture Monitoring 7,

System (primary side leak detection) assures water ingress control.

i I

g Operation of the Main Steam (tube side) portion of the steam generator at a, pressure well above the primary side assures that a tubing failure or slight tube leakage does not result g in the release of primary system containments to the secondary side. It should be noted that the re.9at bundle is at a b.

sliQtly lower pressure en the secondary side. The reheat g system is monitored for primary system leakage.

A-28

O o

j Operating history of the steam generators has resulted i,-

in only one tube failure.

?,

Heavy wall tubing minimizes the possibility of failures due j to erosion or fretting wear.

I I

I 5

Primary side helium purity limits chemical corrosion to 1

1~ the secondary tubeside.

I

d. Based on the above, Steam Generator tubing inspecticn is not O

considered necessary to assure the long term heat removal j, and water ingress protection of the primary system.

I!

D It is,. however, recomended that data be supplied to the NRC which describes material tests performed to date to assure

, the long tenn thermal performance of the bi-metallic welds 9 at the top of the steam generators.

O

.}

O m

A-29

-.~ . _ . - _

A r tACl4 MEN T 4. To P-34.o'2.S (ExcE:2.9tf Frzom P--? '2.00.: l)

.._ )

! pdUe Se.vice Company & Cdende 5909 East 38th Avenue, Denver, Colorado, 80207 Q \ .*

g

.. 1d March 29, 1982 Fort St. Vrain Unit No.1 P-82061 rJv'-63 Mr. Robert Clark, Chief Operating Reactors Branch 3 Division of Licensing Office of Nuclear Reactor Regulatig.i Nuclear Regulatory Commission Washington, D.C. 20553 Docket No. 50-267

Subject:

Fort St. Vrain Inservice Inspection and Testing 1

2 0

References:

(1) Los Alamos Letter

} Q-13:82:5, G-82014 Charles A. Anderson to Robert Clark, dtd January 5,1982

2) P-80014

3) P-80034

4) P-80064 (5) P-80218 Gentlemen:

! PSC has received the above referenced report reviewing the Public Service Company of Colorado's proposed Inservice Inspection and Testing Program, prepared for the Nuclear Regulatory Ccmmission by i

Los Alamos National Laboratory (LANL) and their consultant, Advanced l Science and Technology Associates (ASTA).

It appears that LANL and ASTA generally agree with the modified program proposed by PSC in their submittals for priority category I systems and components. There are, however, instances where the report concluded that specific infomation be provided to the NRC, or that PSC further investigate the applicability of additional examinations outlined in the report.

3

C. o P

@*h

-- ~ ., . - - . .. - _ - . .

.s is **

P-82061 Page 2 i March 29, 1982 i

i O

! Enclosed are PSC's responses to the recommendations contained in

, the LANL/ASTA report. PSC is now awaiting an official NRC response or comments on the referenced Category I submittals, the LANL Review

report and this response to the LANL report in determining the adequacy and acceptability of the approach being used to develop the FSV ISI requirements. PSC is also prepared to meet with the NRC staff to discuss the proposed ISI program and the responses to the LANL/ASTA review as mutually agreeable. Please direct any questions you may have to Mr. Mike Holmes, (303) 571-6711.

Very truly yours, p u.w w 3-H. L. Brey, Manager Nuclear Engineering Division HLB /MAJ:pa Enclosure cc: C.A. Anderson, LANL O

l O

i

.e . .. . .

e t~

I

!O

?

4 PUBLIC SERVICE COMPANY OF COLORADO 4 .

FORT ST. VRAIN NUCLEAR GENERATING STATION i TECHNICAL SPECIFICATION SURVEILLANCE PROGR.LM

() .

i J

t RESPONSE TO THE RECOMMENDATIONS OF LOS ALAMOS NATIONAL LABORATORY Report 0-13:82:5 (j February 19, 1982

<.,,,. o. ,

()

i

)

i SECTION A COMMEN*S RELATIVE TO THE STEAM GENERATOR TUBING Recommendation -

The Report concludes that Steam Generator tubing inspection for Fort St. Vrain is not necessary to assure the long term heat removal and water ingress protection of the primary system. The Report does, however, recommend that data be supplied to the NRC which describe material tests performed to date to assure the long term thermal performance of the bi-metallic welds located at the top of the steam generators.

PSC Resconse -

In describing the steam generator tubing configura-i tion, the Report states that helical tubing in the main section is subheadered and connected to the superheat

() section at the top through a series of bimetallic welds. PSC notes that the tubing in the crossover region between the EESHI bundle and SHII bundle is not subheadered and each tube has one bimetallic weld in the vertical portion of the crossover region. The Report also states that the superheat section is located inside the helical bundle. PSC notes that both the EESHI bundle and the SHII sections are helical bundles and SHII is located above EISHI and not inside as stated.

The Report further states that bimetallic welds are located in the top of the steam generators between the main and superheat sections which are subject to severe thermal conditions and that one failure has occurred in this region. PSC notes that the bimetallic welds are located in a straight vertical section of the crossever region between EISHI outlet and SHII inlet, while the tube failure which occurred was determined to be located in an outlet tube at the botten of a SHII bundle (subheader F of mcdule 3-1-1, as identified in PSC letter P-78007' dated January 13, 1978).

With respecs to the Reporm recuess that PSC provide the NRC with =aterial tes dans relative to the :: ss-(j over bi-netallic welds, PSC notes that such data are included in Appendix A of the FSAR (section A.l.12.3, page 25

,,,.n'.'.'-

O i

I A. l .12.14, A. l.12.16, A. l .12.17, A.1.12. 21) .

Additional information concerning the design and analyses of these welds, as well as an e'aluation of the welds in the Fort St. Vrain confjat ation and environment,was provided by PSC letts P-77084 dated March 7, 1977 to answer NRC concerns which were raised by bi-metallic weld failures of a design used in fossil-fired boilers. The differences in bi-metallic weld design between Fort St. Vrain and fossil-fired boilers were such that no specific areas of concern were identified.

PSC is not aware of any new test data that might be made available to the NRC and considers that no further investigation is warranted. -

It should be noted that the structural integrity of the

. cross-over region in general, which includes the referenced bi-metallic welds, is further assured by

. operational temperature limitations which are more conservative than those established from ASME Code j material characteristics. Correlations have been es-({} tablished between the crossover region temperature and I measured plant operating parameters, to include in I

particular the main steam tempe_ature which is con-trolled as a function of reactor power to prevent the crossover temperature limit from being exceeded. PSC considers that there is currently no technical justi-fication for concerns about the crossover bi-metallic welds.

O page 26

7- -

, ATTAcumsraT 5 TO P-94CC E HELIUM It!LET -

C 3

~

REHEATER l r ALLOY 800 fr _

(SANICRO 31) ,

o LOCATI0fl 0F TUBE LEAKS n , flodule B-1-16" above bottom of SH 2 flodule B-2-3 4" above botton of SH 2 SUPERHEATER 2 [ ,

s

=j j BI-METALLIC

, WELD y _ _

Bottom of SH 2 ,

C ' WJ SUPERHEATER i lg p o

@ ll

• 2 1/4 Cr-1 Mo STEEL ll EVAPORATOR ECONOMlZ ER

@ ll

@ ll C  ;

- , HELIUM OUTLET F (typical)

Gecouaurse. susser,oeR Q%

o d * %o 0

% STEAM Su o> t#- D S C.

Pf2i M AR.3 ' Lc 5ete i _ _1 t i Feedwater , Main Steam Inlet Reheater Outlet Steam Figure 1 Steam Generator Arrangement 25 L

r-ATT Act4M FMT G T'o F-E40 M

-sr. sit.. 0 4 vPeta stu 67 le e.g.

= '

g4e - LIFile6 test

.i,04n 3i 7 9 .g e i

musum mm aspetatte- Q,

[h -(, 100l*F 609 p5 4 660 FT I l.

,I en toets L.14 FT tatn y g 476*P I . h tta 0.3 a0140"mut Q

• l Ik *._' 686 Plea ns.fe.Cr. Ga t e 1 I

.. g y g g .,

Sne0US l SU7tertatte s e -

g ,

h mor pg g a 490 f t U gl hym:

56 i- fusts L.-40 ei .0,*0.C,a, sa . .

0-f t teos

=6 ,

,J' H e 10*F 3 0.e4 ,,, a h* 1 P n 1193*F

$geteestatte t- I I hw 2600 st3 56 fv8tl L t?$ at tacu i

[' 25,e,t y -

la 0e 1e 3 114 C,s. 0.125" t see TO O tol* mall  %

III Cr lit sie

{,

{ tumfoe f 4?

- +,.

l RC0=0mitte 4

1. '7c00s e

.- L0tsta Stu ,

,0 taa 0, ,0.t,

=

b. N? :::: V 'n'".a I .e STtape tiae out

i. ,

IS Tutt$

Me f'fLti 76.' .a

.. .. %Fl(Owaf i tvets tt

i. %

b penseas? CL0 sung e 3 FT. 1/6 s e .

~~

< 4 YOP OF Pult

/ P' t

I I

Liegt ,

e.

h, ' l I d

g ,

l <

4

• """*"""* I se.

Pfeffeafl0e LPete

' **d (40 t e. I .S. ) , ,

i el af. 4 is

} 8' CDeCatit SmitLO

> > . ,6 . e 0.. j fuSt tue0Lt I 3

p. e .

y stetteafl0e testa ef SECDecaay CL0$unt m .

l mom.C30f Paaf or ga('v)tGast

• E Testamat Skifvtl r Class t .q ! l n4 J ,,fsra a CTOs

1. 1. 'l l w stLLows CometCTion

, [ p .M{ g CLa55 s g"a' ($u" \, y / .ais set  ;

llg g*. .s_ outlet 63 *r - me 4 ge . iCos a I 303 psia '.u $ -  ; ;

/, , i _y tl'a essa , a

'S 9

re towarta insgf D ,.,,

i ' ,,

SYt aa O *t 80wse Ps'i oG e

'# ~

Ik# U"U I '

Ftt0wavea part

• ' # l 80wta PiPenG#yv . -

etLLows C3em(Cfice I

{ ,

l se0T atutet piot eon CODE *M I I

$, gg 11 FT I re i n 9

A.

COLO avatar a 4 f;

stuo s C0enCrice-C6 ass a t _ _/

w.. a0 n ta, ,ett, C [

- -- - - - 67 3.,

[ ;? ,

C3tB stataf 8t*t

• 66 3 'S e a

*e [ 'O ' U.. e. Hy"3 ae f .

t ~

c.0 1a ..t.av . eC . n i;

OF y at etssong utC0= case ae0 Coe,a.e.e,.

• • imaa cc0L. a,s aet ;

~-.#,,,,,,,,,,,,,,,t, c eco eg , a 10c t*r p =0? Oratevista e e0iCatt.0 aat SE au Ct ivae ceos C,s Ch.all.a.a.CDs e n oc acet,etg , _ . _____ em a t,.t a, ,, ,t

p. Su0'Catte sat mos C004 C3sacetets pewte pipios 5

T , ,,?~* TQ

[

, . .s  :. .r:h -

( ,

,q !'- ,).

y, 1-s ,  ; j 3 .

, llq - -b ,q':xzt 'F_ ,, . d i i

'g{

I

• Wi W

l,

,: e, 9l im _ g;m77' j i $

.y .y . .. .

3

~ " Q 1 . g .;%

j

_* I u-3

• s. w, ..

rg s

• f 4 4P y (0!: y 0

.  ? 's i 1

.] .<_ ji j ha' to Fig. 14. Steam Generator Steam Section o i Being Assembled to Primary Closure N

_ _ _ _ _ _ _ ._ . ..- - - - _ _ _ . . ._