ML20236U576

Safety Evaluation Re Helium Circulator S/N C-2101 Damage. Util Corrective Action Program Initiated

ML20236U576
Person / Time
Site:	Fort Saint Vrain
Issue date:	11/20/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20236U568	List: ML20236U565 ML20236U576
References
TAC-65992, NUDOCS 8712030171
Download: ML20236U576 (5)

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1. g UNITED STATES

! . NUCLEAR REGULATORY COMMISSION I l wasmoTow. o. c. 2osss .

Enclosure SAFETY EVALUATION BY THE OFF_IC_E OF_ NUCLEAR REACTOR _ REGULATION ON THE HELIUM CIRCULATOR S/N C2101 DAMAGE FORT ST. YRAIN NUCLEAR GENERATING STATION PUBLIC SERVICE COMPANY OF COLORADO n

,L,ICENSE DPR-34. 00CXET NO.: .50-267  ;

INTR 000CTI_0N At a meeting on September 11, 1987 at Recion !Y headquarters, the' Public  :

Service Company (PSC) of Colorado (the licensee) presented a preliminary report on the damage of the helium circulater S/N C-2101 at the Fort St. Yrain j

, Nuclear Generating Station, including a licensee assessment and . justification i l for returning to power. PSC removed the circulator from the reactor on 1 l

July 31, 1987, after indication of excessive speed, shaft wobble .and excess interspace helium leakage. The damaged circulator was replaced with a refurbished spare C-2104 during August, 1987.

The Fort St. Vrain Nuclear Generating Station is a high temperature gas cooled q reactor (HTGR) using helium as the primary coolant. The reactor has two coolant loops, each containing two helium circulators. The circulators y l consist of a single stage helium compressor on a water supported shaf t driven by either a steam turbine or.a Pelton wheel. Labyrinth setls separate

, the bearing water from the reactor coolant on the one side and from the steam .

(orwater)driveontheother.

The "C" circulator in loop 2 started ramping up in speed on July 22, 1987, while the reactor was at 70 per cent power, due to a problem in the control.

system. This caused the "C" circulator to trip. The "D" circulator then ramped uo to compensate for the~ loss of coolant ficw. Reactor power was reduced. Subsequently, tne 'D" circulator tripped from 9200 rpm, anc reac:ce power was reduced to 2 per cent. On returning Loop 2.to service and-increasing the power to the reactor the following day, erratic speed indication ind. )

excessively nign. wobble were observed in the "D" circulator. The decision-  ;

was then made to remove the "D" circulator from operation, and continue to raise the reacter to power on the remaining three Oir:ulators. '

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On July 27, 1987, a high pressure of purified helium was observed in the turbine water drain tank. The most likely source was attributed to a penetration interspace leak through the water-turbine piping in the "D" circulator. The leakage rate was calculated to be 915 pounds per day {

compared to Technical Specification LCO 4.2.9 allowable of 400 pounds per i day. Shutdown of the reactor was then initiated. )1 On July 31, 1987, the decision was made to remove "D" circulator (S/N C-2101) from the reactor for inspection and replace it with an available refurbished I spare (C-2104). The "D" circulator was considered inoperable because (a) interspace helium leakage in excess of the Technical Specifications allowable, and (b) wobble indication greater than permitted by the circulator j operation and maintenance manual.

• The "D" circulator was removed from the i reactor penetration and sent to GA Technologies, Inc., for disassembly on August 14, 1987. ]

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l INSPECTION AND FINDING l Failed circulator hardware pieces were found on the top surface of the weld neck flange of the steam inlet and water piping assembly in the steam outlet I area. The pieces were identified as being from at least three stationary parts which were originally located above the steam turbine rotor:

1) insulation cover, 2) labyrinth seal, and 3) labyrinth spacer. In addition, a piece of lockwire and part of a bolt were found. Chemical analyses of the ,

larger pieces shewed that they were Type 430 ferritic stainless steel, i consistent with the specification for the stationary parts. The Pelton wheel and steam turbine stator were examined after removal of the steam inlet and i water piping assemoly. Nothing unusual was observed on the Pelton wheel anc ]

steam turoine stator. I l

Preliminary metallurgical examination was made on parts recovered from the failed circulator. The components examined were pressure tap bolts, steam ducting to bearing assembly bolts, labyrinth seal mounting bolts, spring plunger, insulation cover, labyrinth seal and labyrinth spacer. The metallurgical examination is continuing. However, the preliminary conclusions ir.dicate that the primary cause of component failure was stress corrosion cracking. EDAX analyses failed to identify a corrosive agent that might initiate and/or propagate the failure mechanism. The surface of the cracked components was oxidized. During assembly of the circulator, molybdenum disulfide was used to lubricate the bolt threads, It is known that molybdenum disuifide may demically met:: vitn :n acueous atmos; hen to 'orm ictive corrocants, sucn as, hycrocen wifide 2nd sulfureous ecid. Oracks acceared to have initiated it stress concentrators; for examole, at the roots of the l

l threacs or :orrosion pit:. 1 ;recracy.ec area initiated ty fat 1gue was found on I

one of the bolts examined.

.. - .. ' ,3, Stress corrosion cracking may be defined as a localized electrochemical attack that occurs along narrow paths when both a corrosive agent and stress are present. For the AMS 5737 Type A286 bolts, a precipitation hardening austenitic stainless steel, the yield and tensile strengths are 95 and 140 lisi, respectively. The licensee stated that the actual tensile strength was 240 ksi for these bolts. In order to avoid stress corrosion cracking in high strength materials, the preload stress should not exceed the yield strength.

Paragraph NF-9729 of Section III of the ASME Code gives a lower limit for preload, but not an upper limit; the maximum preload should be controlled so that the yield strength is not exceeded. The licensee stated that the calculated stresses based on component assembly. torque values were 117 and 139 ksi, respectively, for the solid and hollow bolts. These values are above the yield strength of Type A286 materials. However, these values may be considered suspect because of the use of a calibrated wrench to indicate stress is inaccurate and the actual preload may exceed the allowable by a considerable ira rgin. Net rotation is another quantity that the licensee attempted to relate tu preload stress. The licensee applied four turns rather than two turns to prevent leakage. This method also may be considered suspect. The amount of stretch of the bolt durino tensionina 1s a more direct measure of bolt preload. Although the licensee did not make stretch measurements durina circulator assembly, ultrasonics and strain cauce methods, and load indicatina '

washers are available to more accurately indicate the load.

The materials used for components under metallurgical evaluation by the licensee are as follows:

Component Material Insulation cover AISI Type 430 stainless steel 1/4 - 20 bolts A193 B6 (Type 410 stainless) 3/4 - 16 bolts A296 (AMS 5737) stainless steel Labyrinth seal AISI Type 430 stainless steel Backing Plate AISI Type 430 stainless steel Labyrinth spacer AISI Type 430 stainless steel Spring plunger Carbon steel Lockwasher AISI Type 430 stainless steel Lockwire Inconel 600 The criteria used for the selection of the materials for the circulator components were that they should possess a) adequate mechanical properties, b) ability to survive " hot soak" conditions, c) adequate corrosion resistance, and d) resistance to stress corrosion cracking under operating conditions.

The materials listed meet the design criteria, and were used after heat treatment recommended to minimize the risk of stress corrosion in high t.emperature steam anaer oxidizing ;cncitions.

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CONCLUSION l l

The basic cause for the stre;s corrosion cracking observed in the S/N C-2101 i damaged helium circulator was not identified. The likely corrodant was the reaction product between high temperature oxidizing reheat steam and the lubricant, molybdenum disulfide, used on the bolt threads for circulator component assembly. The stress used for tightening the bolts was not accurately measured and may have exceeded the yield strength of the bolting material. The problem of stress corrosion cracking is generic and not limited -

to the damaged circulator. All the helium circulators in the Fort St. Vrain reactor may be subject to an identical failure. However, there are twelve 1/4-in. bolts that hold the labyrinth steam seal assembly above the turbine wheel and fifteen 3/4-in. bolts that hold the steam ducting to the bearing assembly. Multiple bolt failure would be required to result in similar helium circul.ator damage.

There has been no significant performance anomalies associated with the operation of the currently installed helium circulators to indicate component failure.

One of the first indication would be a reduction in the propensity to self-turbine. The discharge pipe contain strainers and is configured so that ejected parts could not adversely affect the operation of downstream equipment. The failure of 5/N C-2101 helium circulator has emphasized the importance of a comprehensive program for monitoring, inspection and mainten-ance of the helium circulators. The procram is being developed by the l licensee.

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SUMMARY

OF COMPITMENT Until the licensee has completed his evaluation of the damage in S/N C-210.1 helium circulator and until the corrective actions have been implemented, the licensee cormits to restrict Fort St. Vrain operation as follows:

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1. Prior to exceeding 30 percent power, PSC will have demonstrated that all four installed helium circulators can perform their required safety functions. Safe shutdown cooling capability of each circulator will have l been demonstrated by restarting after a 90 minute shutdown and providing l

the equivalent of 3.8 percent primary coolant flow, while driving the Pelton wheels with simulated boosted firewater. (If the lower pressure unboosted condensate provides sufficient flow, this may be used to demonstrate boosted firewater capability). Appendix R cooling capability will have been similarly demonstrated, using condensate Pelton wheel supply for Train A and simulated boosted firewater (or unboosted  ;

condensate) for Train B. Depressurized cooling capability will have oeen demonstrated ;y verifying snaft woobie less than an equivalent of 0.5 mils at 5000 rpm; since the lower pressure water turoine crive tests verify flow and speed characteristics, verifying the stability of the machine assures ac:eptable performance. The operability of helium circulator auxiliaries will be maintained per current Technical Specification requirements.

2. If another helium circulator is found to have failures similar to C-2101, the reactor will not be operated until corrective actions are completed.

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3. PSC will not operate FSV at power levels above 35 percent unless all four-circulators are operating. After a circulator trip, power will be reduced to less than 35 percent in an orderly manner, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, until the reason for the circulator trip is determined, any corrective, actions are completed, and the circulator is returned to operation. ,

4. PSC will implement an enhanced monitoring program that will support circulator performance trend analysis. This program will monitor various ,

performance indicators such as shaft wobble, bearing cartridge j differential pressure, and self turbining capability. In the interim q period until an improved wobble monitoring system is available, PSC will use oscilloscopes to monitor wobble daily or after unplanned speed changes. This interim wobble instrumentation is currently installed.

In addition to the above operating restrictions, the licensee commits to the p following actions: ,

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1. PSC will continue to investigate the failures observed on circulator C.2101 and will provide a final engineering evaluation by January 8,1988. /

This evaluaticr. will identify all corrective actions that PSC considers appropriate and will provide a plan and schedule for their implementation. ,

A meeting with the NRC will be scheduled to discuss the proposed actions. j i

2. PSC will propose a change to Technical Specification SR 5.2.18, to I include an enhanced circulator inservice inspection program. This f l program will be based on the results of the above engineering evaluation. 1

3. PSC will implement the corrective actions identified in the above k :

engineering evaluation, on all circulators. Materials replacements will k be performed as soon as the materials and a refurbished spare circulator are available, approximately the spring of 1988. Other actions will be implemented in accordance with an agreed schedule.

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