ML20235L963

Submits Addl Info Re Atws,Per 880714 Telcon & 881118 Request.Reserve Shutdown Sys & Controls Classified as safety-related & Built Per IEEE-279

ML20235L963
Person / Time
Site:	Fort Saint Vrain
Issue date:	02/17/1989
From:	Brey H PUBLIC SERVICE CO. OF COLORADO
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
P-89045, TAC-59097, NUDOCS 8902280064
Download: ML20235L963 (21)

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$. Public 2420 W. 26thService-Avenue, Suite 100D, Denver, Colorado 80211 $ nIcIsa20b oso February 17, 1989 Fort St. Vrain Unit No. 1 P-89045 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Docket No. 50-267

SUBJECT:

ATWS - Request for Additional Information

REFERENCE:

1) NRC Letter, Heitner to Williams, Dated '

November 18, 1988 (G-88471)

Gentlemen:

A seiephone conference between Public Service Company of Colorado (PSC) personnel and the NRC staff was held on July 14, 1988 to discuss design features of the Fort St. Vrain Nuclear Generating Stat 6n (FSV) which would be expected to mitigate an Anticipated TraiK1ent Without Scram (ATWS) event. At the conclusion of the ,

confeNnce, the NRC staff informed PSC that a formal request would be i submitted addressing their questions on specific FSV ATWS requirements.

In the above referenced letter, the NRC staff submitted their formal ,

request for additional information on the comparison of FSV systems which could mitigate an ATWS event against design guidance for LWR systems which mitigate an ATWS event.

PSC hereby submits, in the enclosed attachments, the requested information.

8902280064 BYU21/

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DR ADOCK 0500 /g7 Ilt -

[ 4 P-89045 Page 2 February 17, 1989 Should you have any questions on this matter, please contact Mr. M.

H. Holmes at (303) 480-6960.

Very truly yours, W 8'"1 H. L. Brey, Manager Nuclear Licensing and Resource Management HLB /JS:emm Attachments cc: Regional Administrator, Region _IV Attn: Mr. T. F. Westerman, Chief Projects Section 8 Mr.. Robert Farrell Senior Resident Inspector Fort St. Vrain-

i Attachment I to P-89045 Page 1 Comparison ' of FSV 's Reserve Shutdown System with Design Guidance Published in the June 26, 1984 Federal Register

.The NRC cover.-letter to the enclosure addressing the Request for Additional Information- also requested that the Reserve Shutdown System (RSS) and any mitigating systems be compared to each line item

' identified in the Table entitled, " Guidance Regarding System and Equipment Specifications" as published in the Federal Register (Vol.

49, No.-124) dated June 26, 1984. Justification for each non-compliance was also requested.

PSC has been unable to readily identify FSV mitigating systems for

- ATWS events which would have to be relied upon to provide protection equivalent to that provided by the Light Water Reactor mitigating systems.(f.e. recirculation pump trip and automatic actuation of the standby liquid control system for BWRs, and auxiliary feedwater actuation and turbine trip for PWRs).

PSC 'has performed a general review of the ATWS rule requirements and the list of operational transients presented. in Attachment 2 to identify any required automatically initiated mitigating systems.

From this review, the identification of FSV mitigating systems for t

ATWS events which would have to be' relied upon to provide protection equivalent to that provided by Light Water Reactor ATWS mitigating systems was not readily apparent. Existing FSAR analyses and

- operational experience demonstrate that sufficient time is available for the _ operators to initiate any required manual actions for event mitigation. -Therefore, PSC has not included any mitigating system comparisons to the Design Guidance items.

The following are the Design Guidance Line items with PSC's response for the FSV Reserve Shutdown System:

1. SAFETY RELATED (IEEE - 279)

Staff Position - Not required but the implementation must be such that the existing protection system continues to meet all applicable safety related criteria.

PSC Response:

The RSS and its controls are classified as safety related and are built in accordance with IEEE-279, August 1968.

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1, Attachment 1 to P-89045 Page 2

2. REDUNDANCY Staff Position -

Not required.

PSC Response:

The RSS consists of two identical independent subsystems, one servicing a group of seven (7) hoppers and the second servicing the remaining thirty (30) hoppers. Each hopper has its own independent gas supply circuit. .Each gas supply circuit utilizes two remotely actuated valves arranged in parallel to provide a redundant valve flow path for the gas (helium) to enter the hopper in the event of a valve failure. The control circuit for each hopper subsystem

contains separate handswitches and control relays for actuation. of the seven or thirty hoppers (see Figure 1). The handswitches for each hopper subsystem control circuit receive AC control power to separate and isolated handswitch contacts from both safety related instrument buses (see Design Guidance 9). This assures that in the event an instrument bus is lost, the valves can still be actuated.

The RSS, 'in addition to the above remote manual actuation from the Control Room, can also be manually actuated locally using either of two different features. One feature utilizes push buttons at the local control panels to initiate valve opening while the second feature utilizes connection of a flexible hose to the valve actuator instrument air line via a nitrogen bottle. This latter feature will allow . opening the gas circuit paths in the event of a loss of instrument air and/or loss of AC power.

The reactivity worth'of t b seven (7) hopper subsystem in itself will compensate for the worst case reactivity accident (uncontrolled withdrawal of the maximum-worth control rod pair). Only in the event  !

that all regular means for shutting down the reactor have been ineffective will the remaining thirty (30) hcpper subsystem be actuated. (See FSAR Section 3.8.3.3)

3. DIVERSITY FROM EXISTING REACTOR TRIP SYSTEM Staff Position - Equipment diversity to the extent reasonable and practicable to minimize the potential for common cause failure is required from the sensors to and including the components used to interrupt control rod power. Circuit breakers from different manufacturers alone is not sufficient to provide the required diversity for interruption of control rod power. The sensors need not be of a diverse design or manufacturer. Existing protection

Attachment I to P-89045 Page 3 l

system instrument-sensing lines may be used.

Sensors and instrument-sensing lines should be selected such that- adverse interactions with existing control' systems are avoided.

'PSC Response:-

The normal reactor trip system, a part of th'e overall Plant Protective System (PPS), functions automatically to protect the core and., equipment from exceeding safety limits by the insertion of all the control rods (Scram).

The Scram circuitry of the PPS consists of'three independent and redundant channels utilizing various sensing circuits (neutron flux, temperatures, pressures, AC power, etc.). Scram actuation is achieved when two-out-of-the-three channels trip. The Scram actuation. causes the opening of a redundant set of Scram contactors which interrupts control rod power and allows control rod insertion.

A typical single Scram channel and contactor arrangement is shown in Figure 2.

The RSS (Diverse Reactor Trip System) is diverse from the normal reactor trip system (Scram) to the following extent:

The RSS is completely manually actuated under procedural control with no design capabilities for automatic initiation.

The RSS utilizes no integral sensing circuits for actuation. The PPS neutron flux sensing circuits are used for verifying

-subcriticality after RSS insertion.

The RSS utilizes gas pressure to rupture a disc to allow insertion of the negative reactivity material ratier than interruption of AC/DC power for the control rods.

l

• The RSS is not an integral part of the PPS.

The only commonality between the RSS and the Control Rods is that they share the same overall Control Rod Drive and Orificing Assembly (CRD0A) housing for each region.

Sufficient negative reactivity is provided to shutdown the reactor to refueling temperatures from any. ope. rating condition (without movement of the control rods) even if one RSS hopper of the 37 hoppers was inoperable (such as gross failure of a CRDOA).

Attachment I to P-89045 b

Page 4 4 ELECTRICAL INDEPENDENCE FROM EXISTING REACTOR TRIP SYSTEM Staff Position.- Required from sensor output to the final actuation device at which point non-safety related circuits

-must be isolated from safety related circuits.

PSC Response:

The RSS and the- overall plant protective system are all safety related and are designed and built to IEEE-279, August 1968. Both the RSS and PPS Scram actuation circuits are totally independent electrically from one another.

The test circuits are electrically isolated from the actuation ci rcui ts.

See Design Guidance 9 for discussion of power supplies.

5. PHYSICAL SEPARATION FROM EXISTING REACTOR TRIP SYSTEM Staff Position - Not required, unless redundant divisions and channels in the existing reactor trip system are not physically separated. The implementation must be such that separation criteria applied to the existing protection system are not violated.

PSC Response:

The RSS and the PPS Scram system are physically separated and their design is consistent with the physical separation / segregation design criteria for bus and loop designation of plant safety systems.

One instance where physical separation is not achieved is in the use of a multi-conductor cable (24/c) which provides power to the CRD0A from the control rod drive motor control center, (See Figure 3). Two of the. twenty four conductors are used to provide power to the pressure switches used in the hopper test circuitry. Any postulated failure in this cable will not prevent actuation and release of the hopper material.

As discussed in Design Guidance 3, a physical commonality exists in that the Control Rods and the RSS are housed in the common Control Rod Drive and Orificing Assembly (CRDCA). Operation of either system is totally independent of the other system. (See Figure 4).

l Attachment 1 to P-89045 Page 5

6. ENVIRONMENTAL QUALIFICATION Staff Position - For anticipated operational occurrences only, not for accidents.

PSC Response:

The RSS control circuits are environmentally qualified to 10 CFR 50.49 requirements. The RSS hoppers and rupture discs are qualified for the in-core environment to which they are exposed.

7. SEISMIC QUALIFICATION Staff Position - Not required.

PSC Response:

The RSS is seismically qualified to function following a seismic event.

8. QUALITY ASSURANCE FOR TEST, MAINTENANCE, AND SURVEILLANCE Staff Position - The Commission has released a Generic Letter (85-06, April 16, 1986) in which is provided the explicit Quality Assurance (QA) guidance required by 10 CFR 50.62. While Appendix B is viewed as a useful reference in which to frame the staff's guidance for non-safety related ATWS equipment, it does not meet the intent of the ATWS QA program.

The equipment encompassed by 10 CFR 50.62 is not required to be safety related, therefore, less stringent QA guidance is acceptable. This letter incorporates a lesser degree of stringency by eliminating requirements for involving parties outside the normal line organization and requirements for a formalized program and detailed record keeping for all quality practices.

PSC Response:

The RSS is classified as safety-related, and is governed by the FSV  ;

Technical Specifications and associated Surveillance Requirements to l demonstrate operability. Therefore, the RSS is encompassed within '

the existing For6 St. Vrain Appendix B Quality Assurance Program.

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l Attachment 1 to P-89045 Page 6 l- 9. SAFETY-RELATED (IE) POWER SUPPLY Staff Position - Not required, but must be capable of performing j safety functions with loss of offsite power. -

Logic power must be from an instrument power supply independent from the power supplies for the existing reactor trip system. Existing RTS sensor and instrument channel power supplies may be used provided the possibility of common mode failure is prevented.

PSC Response:

The Scram system utilizes power from both the safety related 120 VAC l Instrument Buses and the 480 V Motor Control Centers.

Operation of the control rods (insertion into the core) is dependent on the interruption of DC power (rectified from the AC) to the brake motor.

As discussed in Design Guidance 3, the RSS neither uses sensors nor

. instrument channels for actuation. Therefore, the RSS is not dependent on any channel or logic power. The RSS does util.ize the same safety related 120 VAC Instrument Buses that power the Scram system to supply control power to'the gas supply circuitry. These buses are non-interruptible power supplies which can be powered from the onsite sources and as such loss of offsite power will not preclude RSS operation. Each of the two handswitches associated with each subsystem is powered from both instrument buses to separate and isolated contacts. The failure of any one instrument bus will not preclude . operation of the entire RSS. Failure of both instrument buses would preclude remote RSS actuation but manual actuation locally via the ' flexible hose connection would still be available.

(See Design Guidance 2).

10. TESTABILITY AT POWER Staff Position - Required PSC Rersonse:

The reliability of the RSS to perform its reactor shutdown function is maintained by periodic in situ tests and inspections during operation. These verification of operability tests is implemer.ted per the existing FSV Technical Specification surveillance. Offline rupture disc tests are also performed. The operation of the valves within the control circuit can also be individually tested at power.

.These testing capabilities were evaluated by the NRC staff and found

i Attachment 1 to P-89045 Page 7 L.

acceptable as documented in a Safety Evaluation, by NP.C letter dated March 12, 1987 (G-87074).

11. INADVERTENT ACTUATION b Staff Position - The design should :be such that the frequency of inadvertent reactor trip and challenges' to other r safety systems is. minimized.-

l PSC Response:

The frequency. of inadvertent actuation of the RSS is considered low i since the system. is neither actuated automatically nor utilizes any sensing circuits Lfor. initiation. Deliberate operator. action is required to initiate the seven (7) hopper and thirty (30) hopper subsystems by manipulating the appropriate two handswitches for each

. subsystem. Overpressurizing a hopper (or subsystem) or operator error - during surveillance- testing could result in an inadvertent actuation.

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1 Attachment 2 to P-89045 Page 1 Response to the NRC's Request for Additional Information

- The following is PSC's response to the request for additional information items specifically identified in the enclosure to the NRC letter of November 18, 1988. These items request information associated with the considerations given to all'possible ATWS events at FSV and the manual initiation requirements for the.RSS. j NRC-Request 1 Provide information to confirm to the staff that all possible anticipated operational transients which could occur at Fort St.

Vrain were considered (define each transient and relate to the plant's existing. accident / transient spectrum).

PSC Response 1:

A review of the updated Final Safety Analysis Report and the Emergency Procedures identified the following list of accidents and operational transients -which could possibly occur at the Fort St.

Vrain Nuclear Generating _ Station. Each operational transient is identified by the expected frequency of occurrence (in the time 1 period from January 1, 1989' to June 30,. 1990) .and whether an i automatic reactor Scram. is required to mitigate the transient (Y = i

. yes, N = no).

ATWS~ events are those for which a scram would normally occur and events which could reasonably be anticipated to occur, at least once, over the remaining operating life of FSV. As stated in PSC letter dated 12/5/88, Williams to Calvo (P-88422), FSV will be permanently snut _down on or before June 30, 1990. Based on the following Table, FSC considers that there are only three potential ATWS events at FSV.

These are: 1) steam generator tube failure and/or a steam / water dump, 2) loss of outside electric power followed by a turbine trip and failure to restore power to 2 of 3 essential buses within 35 seconds (failure of both SDG output breakers to close automatically),

and 3) a temporary interruption of forced circulation cooling.

Frequency of Required Potential Operational Transient Occurrence Auto-scram

1. Column Deflection and Misalignment <1 N

2. Fuel Element Malfunctions <1 N 1

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1' Attachment 2 to P-89045 Page 2; 1

' Frequency of Required Potential Operational Transient Occurrence' Auto-scram

3. Hisplaced Fuel Element <1 N

4. Blocking of Coolant Channel <1 N

5. Control Rod Malfunctions <1 N

6. .Or'ifice Malfunctions <1 N-

7. Loss of Core Support Floor Cooling <1 'N.

'8. Helium Circulator Malfunctions 21 N

9. Helium Circulator Auxiliary. System Malfunctions. 21 N

10. Moisture Monitor Failure < 1. N

11. Loss of Access to Central Control Room <1 N

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12. PCRV Prestressed Tendon Failure <1- N

' 13. Failure of Thermal Barrier and. Liner Cooling <1 N

14. Steam Generator or Circulator Penetration Overpressure <1 N

15. Loss of Reactor Generated Steam (HELB) <1 Y

16' . Failure of Main Condenser 21 N

17. Steam Generator Tube Failure-Steam / Water Dump 21- Y

18. Loss of Main Feedwater or Condensate Line <1 N

19. Simultaneous Loss of all Three Boiler Feedpumps (1 Y i

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4 Attachment 2 s to P-89045 Page 3 Frequency of Required Potential Operational Transient Occurrence Auto-scram

20. Low Feedwater Flow 21 N

21. Safe Shutdown Following Earthquake, Tornado or HELB <1 Y

22. Circulating Water or Service Water Piping Failure, Condensate Storage Tank Failure <1 N

23. Loss of Main Generator 21 N

24. Loss of Unit Auxiliary Transformer 21 N

25. Loss of Reserve Auxiliary Transformer 2.1 N

26. Loss of all Outside Electric Power Followed by a Turbine Trip 21 N

27. Loss of all Outside Electric Power Followed by a Turbine Trip and Loss of Standby Diesel Generator >1 N

28. Loss of all outside Electric Power Followed-by a Turbine Trip and Failure to Restore Power to 2/3 Essential Buses within 35 Seconds (Failure of Both SOG output Breakers to Close Automatically) >1 Y

29. Loss of an Instrument Bus or and Inverter 21 N

30. Loss of a DC Bus 21 N

31. Helium Purification System Malfunctions 21 N

32. Helium Storage System Malfunctions <1 N l

33. Nitrogen System Malfunctions <1 N J

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Attachment 2 '

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.. Frequency of Required l Potential Operational Transient Occurrence Auto-scram

34. Heavy Load Handling Malfunctions <1 N l35. -PCRV Relief Valve Opens <1 N

36. ' Loss of all' Hydraulic Pressure in One System <1 N

37. Fire <1 N

38. Loss of an Instrument Air Header <1 N

39. Interruption.of Forced Cooling 21 Y

40. Rod Withdrawal Ac'cidents (Reactivity Accidents) <1 Y

-41. Maximum Credible Accident <1 Y

42. Design Basis Accident -'l- <1 Y

43. Design Basis Accident 2 <1 Y

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44. -Station Blackout.(prolonged period) <1 Y

45. Loss of Normal Shutdown Cooling:

Cooling with'one circ on steam Cooling with one circ on water 21 N

46. Accidents Involving Gas Waste System <1 N

47. Main Turbine Trip >1 N  ;

l 1

Attachment 2 to P-89045 Page 5 NRC Request 2:

'For each . anticipated operational transient considered, confirm the operation of the reserve shutdown system (RSS) would be sufficient to mitigate consequences of the event assuming a failure of the existing reactor. trip system (RTS) to scram the reactor.

PSC Response 2: l The' RSS by itself is capable of safely shutting down the reactor to refueling temperature and maintaining subcriticality for up to two weeks, with the maximum worth hopper unit inoperable (FSAR Section 3.5.3.3). Therefore the RSS has the capability of completely duplicating the function of the control rods in an ATWS event for a sufficient period of time which would permit insertion of additional control rods to augment the RSS.

For the three potential ATWS events defined in Response 1 of this attachment, the .following is an overview of how the operators would mitigate the event.

STEAM GENERATOR TUBE FAILURE - STEAM / WATER DUMP Discussions of steam generator tube leaks in the FSAR are focused or )

design basis events of an offset rupture of a subheader with initial  !

water / steam 1eak rate of 35 lb/sec. These are discussed in FSAR Section 14.5.2.3 and are considered to be emergency / faulted events.

They are, therefore, not a candidate for an ATWS event.

A' small steam generator tube leak of less than or equal to 1 lb/sec is a potential ATWS event for Fort St. Vrain and will be discussed.

See Response 3 of this attachment for a description of the one steam generator tube leak that occurred during power operation at Fort St.

Vra in. In that case, the leak rate was much less than 1 lb/sec and many hours were available for the recovery.

For this discussion it will be assumed that the leak rate is 1 lb/sec. Moisture concentration would reach 500 PPMV in the primary coolant at slightly less than 20 seconds. The Plant Protective System (PPS) Dewpoint Moisture Monitors (DPMMs) would trip most likely within a minute of the onset of the event. The additional delay time is due to the DPMM sample transport time and mirror fogging time. The trip would call for a loop trip and dump of that loop's steam generator to the steam / water dump tank, a feedwater flow runback to -about 27% of rated feedwater flow, and a reactor scram.  ;

For purposes of this discussion, the reactor scram is assumed not to have occurred. Power to flow ratio would exceed 1.0 causing the core

Attachment 2 i to P-89045 Page 6 to heat up. This, in turn, would cause the power to decrease due to the negative temperature coefficient of reactivity. Based upon prior LHTGR ATWS studies, it is anticipated that the power to flow ratios would stay above 1.0 until the reactor is made subcritical either by the insertion of the control rods or the RSS.

The worst case event would be if the intact steam generator loop were  !

dumped as opposed to the leaking loop. In this case s team /wa ter would continue to leak into the primary circuit at the rate of 1 lb/sec. Primary coolant pressure would increase as a consequence of the continuing ingress. As discussed in FSAR Section 14.5.2.1, the i contribution of steam pressure and steam-graphite reaction products to the original primary coolant system pressure is essentially linear with the mass of steam added and the fraction of steam reacted with core graphite. This linear characteristic has a slope of about 2.1 psi per 100 lbs of steam added if there is no s team-graphite reaction, and about 4.2 psi per 100 lbs for completed reaction of the steam with graphite. These characteristics assume constant helium l temperatures in the primary coolant system. At an ingress rate of 1 l lb/see the pressure would increase to the PPS programmed high primary coolant pressure trip of approximately 50 psi above normal PCRV pressure. The time for this to occur would be between 21 and 42 minutes dependent if complete steam graphite reaction is assumed or if no steam graphite reaction is assumed. During this time period of at least 21 minutes, the operators would bring the reactor to a subcritical status by either inserting the control rods or by insertion of the RSS. Once the reactor is placed in a shutdown status prior to the programmed high pressure trip, the small steam generator tube leak would essentially be bounded by large steam generator tube leaks evaluated in FSAR Section 14.5.2.3. No calculations have been made on the quantity of graphite that would be reacted.

The above described sequence for the assumed 1 lb/sec steam generator tube leak rate can also be related to a spurious trip resulting in a i steam / water dump of one steam generator without the scram which normally accompanies a steam / water dump. The sequence of events would be as previously described except there would be no water l i

ingress and no consequent primary coolant pressure rise due to this  !

water ingress. The failure to scram (ATWS assumption) would again 1 result in power to flow mismatch and possibly some increase in PCRV {

pressure due to an increase in primary helium coolant temperature.

Operator insertion of the RSS in 21 minutes or less would terminate the transient comparable to the previously described 1 lb/sec steam generator tube leak event. In this case with no water ingress, the PCRV pressure rise would be limited and there would be no graphite oxidation.

l l

l

' Attachment 2 to P-89045

,. Page 7 INTERRUPTION OF FORCED CIRCULATION COOLING /SHORT TERM STATION  !

BLACK 0UT Both! of these events result in the loss of the helium circulators on their steam drives. Without the normal reactor scram, the reactor

. power is reduced to decay heat power generation rates within minutes solely due to the negative temperature coefficient of reactivity. i See PSC. Response 3 of this attachment for a discussion.of shutdown characteristics with time and the reactor fuel temperature response.

Recovery operations from these events would be those activities j

necessary to shutdown the reactor and reestablish forced circulation l cooling with at .least one steam generator loop and at least one  ;

helium circulator.

In the ATWS event the operators must place the reactor in a scram status' prior.to the initiation of forced cooling or the reactor would again go critical. -By' insertion of the RSS prior to initiation of forced circulation cooling a normal plant recovery would be

. accomplished. Per PSC Response 3 of this attachment at'least 10 minutes would be available to insert the RSS.

An element of the Reactor Scram Emergency Procedure is the immediate response of the operators.to ensure the reactor' is shutdown. This response requires the immediate insertion of the Manual Scram. If the reactor is not in a subcritical condition following the scram input, -the ' operators are directed to insert the 7 hopper RSS subsystem. Should the reactor- still not be subcritical, the remaining 30 hopper RSS subsystem is inserted.

NRC Request 3:

Address the adequacy associated with manual initiation requirements of the RSS as related to all possible postulated anticipated operational transients considered. The information should include a scenario associated with the anticipated operational transient (including related plant operating conditions just prior to the 1

transient) which could result in the worst-case (shortest) time that would be allowed to elapse before the RSS must be activated manually to shut down the reactor in lieu of the failed RTS.

PSC Response 3:

Analyses were performed in 1978 using the BLOOST Code to demonstrate the shutdown characteristics of the Fort St. Vrain reactor with no primary coolant flow and no reactor scram. In this case, the reactor power is reduced solely due to the negative tempera ture coefficient of reactivity. After two minutes, reactor power drops from 842 l

l 1

b

l

]

J Attachment 2 to P-89045

,,e Page 8 l

.~

l MW(th) to 145 MW(th) with no scram as opposed to 29 MW(th) with an automatic scram. -At six minutes, the no scram and scram case have nearly identical decay heat power generation rates of about 25

'MW(th). At ten minutes the decay heat power generation rate is 20 4 MW(th) for the scram and no scram cases. Maximum. fuel temperature computed by the BLOOST Code for the non-scram case increased from an initial 1580 degrees F to 2300 degrees F at 2 minutes. At 10 minutes, which was .the duration of the analysis, the maximum fuel temperature was 2625 degrees F. This is below the FSAR fuel failure temperature limit of 2900 degrees F.

The above is directly appiicable to the interruption of forced circulation cooling and the short term station blackout since the immediate consequence of the latter event is also an interruption of forced circulation cooling. The slow reactor core temperature response- without the normal reactor scram demonstrates the adequacy i associated with manual initiation of the RSS including sufficient time for the operators to perform other functions as may be required.

i Small steam generator tube leaks (less than or equal to 1 7b/sec) which might be anticipated during the remaining planned power operation of Fort St. Vrain and coupled with no reactor scram have not specifically been analyzed. However, the sequence of events and their timing for. the steam generator tube leak that occurred on

. November 30, 1977 indicate there would be ample time for any required operator manual actions. This event is one of two steam generator tube leaks which have occurred.over the lifetime of Fort St. Vrain but is the only one to occur during power operation.

On November 30, 1977 the reactor was operating at a steady power level of approximately 50%. At 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> shift personnel were alerted that a water ingress had occurred or was in progress. There was a slight increase in hydrogen level and a 18.3 PPMV increase in H20 level between 0800 and 1000. At 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> with the H20 level at 52 PPMV, reactor shutdown was initiated. At 1203 hours0.0139 days <br />0.334 hours <br />0.00199 weeks <br />4.577415e-4 months <br /> Loop 2 steam generators were arbitrarily selected and manually dumped. This turned out to be the intact loop. At 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br /> Loop 2 was returned to service, Loop I was isolated and dumped to the steam water dump tank at 1705. Primary coolant was detected in the dump tank at this time. The total moisture ingress was 72 pounds. The maximum moisture ingress rate was estimated to be 43 lb/hr and the resultant hole size to be 0.01 inch in diameter.

A sequence of events which permit 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to commencement of reactor i shutdown to 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> for completion of all required operations, provides more than sufficient time for manu:1 initiation of the RSS should it have been required.

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