DUKE POWER COMPANY POWER DUIDIoo

422 SOUTH CHURCH STREET, CHARLOTTE. N C 2e242

WILLIAM 0. PARKER.JR.

VICE PacsorE.T TCLE.ONE: ASA 704

STE.. PpoocT,0 373-083

August 8, 1975

Mr. Norman C. Moseley, Director U. S. Nuclear Regulatory Commission Suite 818 230 Peachtree Street, Northwest Atlanta, Georgia 30303

Re: Oconee Nuclear Station Docket No. 50-287

Dear Mr. Moseley:

My letter of June 27, 1975, transmitted to you Abnormal Occurrence

Report AO-287/75-7, Excessive Reactor Coolant System Cooldown Rate.

The following information provides additional information relating to

this occurrence and associated corrective action.

As stated in AO-287/75-7, when reactor power level had decreased in the

course of a routine maintenance shutdown, a minor system transient

occurred, which resulted in the opening of the power-operated relief

valve 3RC-66. Prior to the system transient, reactor power was being

reduced from 100% FP to 15% FP in an orderly manner by the Integrated

Control System. When 15% FP was reached, unit load demand was 65 Me, and power generation was 115 MWe. This difference between unit load

demand and power generation existed because the reactor was operating

at its low limit of 15% FP while in automatic ICS control and could not

further follow unit load demand. Meanwhile the control operator placed

the turbine control station in manual, leaving the ICS in the "load

tracking" mode. This led to a rapid increase in unit load demand to

match the generated megawatt output. In the meantime, the main steam

bypass valves opened; and when the main steam pressure decreased, the

valves closed. The ICS control of feedwater flow could not follow the

rapid change in unit load demand and steam pressure; consequently,

feedwater flow and steam generator level oscillated, resulting in the

Reactor Coolant System temperature and pressure transient.

The power operated relief valv,,ARC-66, opened when RCS pressure

reached 2255 psi but failed to\elbse when the pressure dropped below

2220 psi, although the open/cl~sed lights in the control room did not

indicate that the valve was opqp. Consequently, RCS pressure dropped,

the reactor tripped on low pre sure, and the HPI system actuated.

79 05lO 3

Mr. Norman C. Moseley Page 2 August 8, 1975

Although the operator closed the isolation valve, 3RC-4, immediately

after the reactor trip to terminate the depressurization, the valve was

reopened because of the rapidly rising pressurizer level. Valve 3RC-4

was finally closed when RC pressure reached 800 psi, terminating the

pressure transient. The subsequent controlled cooldown of the Reactor

Coolant System, when combined with the temperature drop during the

transient, resulted in a cooldown of 1010 F during the first hour when

temperature was below 5300 F, contrary to the provisions of Technical

Specification 3.1.2.3. The transient and associated events also

caused the quench tank rupture disc to blow open, Mirror insulation

to be separated from the bottom nozzle of the pressurizer, and the

release of approximately 1500 gallons of reactor coolant to the Reactor

Building sump.

The release of reactor coolant did not cause any significant increase

of radiation level in the Reactor Building, and no radioactivity was

released into the environment.

As addressed in AO-287/75-7 , the excessive cooldown rate associated with

the transient has been evaluated and it was determined that the health

and safety of the public was not affected. No other system limits were

exceeded.

The failure of 3RC-66 to close and the malfunctioning of the valve

position indication in the Control Room have been investigated. It

has been found that the valve was stuck in the open position because

of heat expansion, boric acid crystal buildup on the valve lever,

rubbing of the level against the solenoid brackets, and bending of

the solenoid spring bracket. The valve was repaired and reinstalled.

The malfunctioning of the valve position indication was not observed

when the repaired valve was reinstalled. This malfunctioning was

apparently caused by the sticking of the solenoid plunger at slightly

less than the full open position or by the crud buildup around the

plunger operated microswitch to the open/closed lights.

Additionally, to prevent recurrence of this incident, the following

corrective actions have been or will be implemented.

1. The unit shutdown procedures for all Oconee units have been revised

to include a change that will prevent decreasing unit load demand

below 120 lIee before placing the ICS in Lhe tcacking mode. This

would minimize the error between the unit load demand and generated

power and thus would reduce the possibility of feedwater flow and

RCS transients.

2. The Units 1 and 2 power-actuated pressurizer relief valves will be

examined as socn as possible for any indication of boric acid

crystal buildup.

Mr. Norman C. Moseley Page 3 August 8, 1975

3. To verify the proper functioning of RC-66, a test to cycle RC-66

prior to startup with a test signal corresponding to 2285 psi will

be incorporated into the station operating procedures.

4. The quench tank rupture disc has been replaced, and the bottom

nozzles on the pressurizer were dye penetrant tested and the

Mirror insulation replaced.

5. Operating personnel have been advised of this incident with specific

instructions that immediate closure of 3RC-4 is the proper corrective

action for such an occurrence.

Ver truly yours,

William 0. Parker, Jr. -j

PMA:vr

cc: Mr. Angelo Giambusso

bcc: Mr. W. S. Lee Mr. S. P. Hellman

Mr. W. H. Owen Mr. H. B. Tucker

Mr. J. E. Smith (15) Mr. G. A. Olson

Mr. J. W. Cox Mr. M. S. Tuckman

Mr. L. Lewis (10) Mr. T. S. Barr

Mr. J. 0. Barbour Mr. R. 0. Sharpe

Mr. K. S. Canady Mr. W. A. Coley

Mr. D. C. Holt Mr. P. M. Abraham

Mr. M. D. McIntosh 1.4.3.5

Mr. P. H. Barton 1.10.3.3

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ITEM 12

Review your prompt reporting procedures for NRC notification to assure very

early notification of serious events.

Response

The Oconee Nuclear Station Emergency Plan (Station Directive 3.8.5) and the

Administrative Policy Manual for Nuclear Stations (Section 2.8 - Reporting)

include prompt reporting procedures for NRC notification of serious events.

These procedures have been reviewed to assure that the requirements of

10CFR20, 10CFR21, and Oconee Nuclear Station Technical Specification 6.6.2.1.a

(."rompt Notification with Written Followup) are addressed.

Station Directive 3.1.5 (Notification of Station Management) requires prompt

notification of the Operations Unit Coordinator and/or the Station Manager

after any of the following events, among others:

(a) A reactor trip

(b) An unexplained reactivity change

(c) A significant increase in reactor coolant leakage

(d) Loss of offsite power

(e) Major component or equipment failure

(f) Accidental release of liquid or gaseous activity

(g) Known or suspected fuel cladding failure

(h) Suspected initiation of or significant increase in primary-to secondary

leaks

The NRC Resident Inspector and/or the NRC/OIE, Region II, office is to be

notified promptly following determination of any event of regulatory and/or

nuclear safety significance.

DUKE PO- R COMPANY OCONEE UNIT 1

Renort Number: RO-269/782 7

Renort Date: January 15, 1979

Occurrence Date: December 16, 1973

Facility: Oconee Unit 1, Seneca, South Carolina

Identification of Occurrence: ES Actuation Following Reactor Tripo

Conditions Prior to Occurrence: 9P' Full Pcwer

Descri:tion of Occurrence:

A: approxiately 1100 hours cn December 13, 1973, the Reactor Coclan:

System average temperature (Tave) statalarm became erra:ti and a work

recuest to investigate was issued. On December li, 197S during invest

igation of the erratic statalarm behavior a short occurrea cn :ne power

cord feeding the ICS Tave recorder.o The short caused an erroneous 1ow

Tave indizaion of approxi=azaly 13 F. The ICS began wizhdrawing control

rod group 7 to maintain Tave. The reactor tripped on high pressure/temp

erature at 1636 hours. Both normal feedwater pumps tripped on highI

discharge cressure. The emergency.' feedwater pump started, then stop'ped

when the ncrmal feedwater pumps reset and restarted. 3 appro:-:imately

1638 hours :ne level in the OTS's had dropped to 6 andc 0 inches respec

tively (normal level is more than L10 inches). The emergency :eawater

oump started to feed the B generator. Levels in the A ceneracor returned

to normal by 1641 hours but the S generator increased to 35 inches and

then returned to 0. This possibly resulted from malfunctioning of valves

recuired for this particular flow oath. T.-e B genera:ctr as re::iec

thrcugh the emergency feedwater header at 1649 hours. ES Channels 1 and n

(High ?ressure Injection) tripped on low RCS pressure (15C0 psig) as

required by Tecnnical Specifications.

Avoarent Cause of Occurrence:

The ES actuated on low RCS pressure due to ina-ility to maintain feedwa:er

flow to t he OTSG's sufficient to maIntain adequa:e s:ar:uo >eve. .he

feedwa:er problems evident1 resulte: from improper opera:ion of one or

more feedwarer valves (D-AS, -A7) involved in switching o

emer-ency f1cw oaths. The initia:1 grjent, the short in me .averecorders

power cord, was apparently an isola ec,:aJure.

Anal-:s.is of Occurrence:

The heal:h and safety cf the gener 4 uli were not adversely affecte:

as a result of :his occurrence. Sot the RS and ES syscams n

as required.

DUKE P(: ER CCPAN,': OCONEE UNIT i

Pg 2

Corrective Actions:

The power cords that suoply the Tave recorders on Unies 2 and 3 nave been inspected and found to be in accep:able condition.

The potential v malfunctioning valves, FD'- 7, - S will be thoroughly checked during a later outage. Further review of the impac: of zte transient on the OTSG's is also being performed.

N FORM : U. S. NUCLEAR AEGULATCAY CM.misiCN (7-77)

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EVENT OESCRIP1TON ANC PQCaA8LS CZNSECUENCES

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Sinvescigation was iniciated. During the inves:a:icn, on December 1i, 197S

the power cord suopling the Ta:e recorder shorted caus:ng an a-Darent knot

real) drop in Tave of 13 F. As the ICS a-he=cad to ccree: .2ve ts

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ITEM 3

Review the actions required by your operating procedures for coping with transients

and accidents, with particular attention to:

a. Recognition of the possibility of forming voids in the primary coolant

system large enough to compromise the core cooling capability, espe

cially natural circulation capability.

b. Operator action required to prevent the formation of such voids.

c. Operator action required to ensure continued core cooling in the

event that such voids were formed.

Response

The following emergency procedures have been reviewed and revised to include actions

required to cope with primary coolant system voids:

EP/O/A/1800/0 8 Steam Supply System Rupture

EP/O/A/1800/04 Loss of Reactor Coolant

A graph of the properties of water and saturated steam has been added to the above

listed procedures. The unit computer is available to provide the operator with

saturation temperature versus pressure information. Also, a computer program has

been established to allow the operator to read selected incore thermocouples for

core temperature.

The above procedures have been revised to assure the operator is made aware of:

(a) The possibility of void formation by emphasizing the operator's need

to check for saturation or near saturation conditions in the reactor

coolant system.

(b) The action required to prevent void formation by checking reactor

coolant pressure and temperature following reactor trip to ensure

that there is at least 500F subcooling of the reactor coolant.

(c) The requirement to assure core cooling in the event of void formation

by the operation of at least one reactor coolant pump per loop and the

continuation of HPI operation as described in the response to Item 4.

Operations personnel have been instructed on these procedural changes. Licensed

shift personnel will be sent to the Babcock and Wilcox training simulator, which

has been programmed to demonstrate the events of the Three Mile Island event, for

training expeditiously consistent with scheduling constraints.

ITEM 4

Review the actions directed by the operating procedures and training instructions

to ensure that:

a. Operators do not override automatic actions of engineered safety

features.

b. Operating procedures currently, or are revised to, specify that

if the high pressure injection (HPI) system has been automatically

actuated because of low pressure condition, it must remain in opera

tion until either:

(1) Both low pressure injection (LPI) pumps are in operation

and flowing at a rate in excess of 1000 gpm each and the

situation has been stable for 20 minutes, or

(2) The RPI system has been in operation for 20 minutes, and

all hot and cold leg temperatures are at least 50 degrees

below the saturation temperature for the existing RCS pres

sure. If 50 degree subcooling cannot be maintained after

HPI cutoff, the HPI shall be reactivated.

c. Operating procedures currently, or are revised to, specify that in the

event of HPI initiation, with reactor coolant pumps (RCP) operating,

at least one RCP per loop shall remain operating.

d. Operators are provided additional information and instructions to not

rely upon pressurizer level indication alone, but to also examine

pressurizer pressure and other plant parameter indications in evalua

ting plant conditions, e.g., water inventory in the reactor primary

system.

Response

Operator training includes specific instructions regarding the overriding of auto

matic actions of engineered safety (ES) features. The operators have been in

structed to determine whether or not a valid ES actuation has occurred by verify

ing actual reactor building pressure and reactor coolant system pressure instrumen

tation. If these instruments indicate a valid ES actuation, then the operator is

not to prematurely override the automatic actions. However, if the instruments

indicate a spurious actuation, the operators will take appropriate actions in order

to prevent any damage to the plant that might occur due to this inadvertent actua

tion. Formal briefings on the events at TMI-2 have been held with operations

personnel to assure they are familiar with the events of the incident and proper

methods to cope with the transient. Additional formal training on the proper re

sponse to the events at TMI-2 will be conducted with emphasis placed on the oper

ator's ability to recognize the possibility of void formation and the need to pro

vide adequate core cooling.

ITEM 4 (Continued)

The Emergency Procedures listed in the response to Item 3 have

been revised to

include the specific actions of 4b, c, d above.

Depending upon the nature of the low pressure transient, appropriate operator

action may be required to alter the operation of the high pressure injection

system and/or the reactor coolant pumps in order to prevent an unsafe reactor

coolant system condition.

ITEM 5

Verify that emergency feedwater valves are in the open position in accordance

with Item 8 below. Also, review all safety-related valve positions and posi

tioning requirements to assure that valves are positioned (open or closed) in

a manner to ensure the proper operation of engineered safety features. Also

review related procedures, such as those for maintenance and testing, to ensure

that such valves are returned to their correct positions following necessary

manipulations.

Response

Appropriate feedwater valves have been positioned and visually verified to pro

vice two independent steam generator auxiliary feedwater flow paths. A more

detailed explanation of the actual flow paths is provided in response to Item 8.

Operating procedures for the following systems have been reviewed to assure that

valves are positioned in a manner to assure that engineered safeguards (ES)

systems and related equipment can perform their intended functions:

Emergency Feedwater System

High Pressure Injection System Reactor Building Spray System Low Pressure Injection System Reactor Building Cooling System Penetration Room Ventilation System Core Flooding System Valves Associated with Containment Isolation Function

Operating procedure OP/O/A/1102/06, Removal and Restoration of Station Equipment,

establishes the methods to be used to remove station equipment from service or to

change its status as required by another operating procedure. Requirements are

included to provide documentation of sufficient testing to assure that the equip

ment is fully operable prior to returning the equipment to service. This proce

dure also provides steps to identify any equipment that is inoperable, out of

specification or does not otherwise meet necessary operational requirements.

Procedures applicable to maintenance and periodic testing have been reviewed and

verified to include provisions to assure that all components, including valves,

are returned properly to service following maintenance or testing. The following

procedures have been revised:

IP/O/A/0310/7B ES Logic Subsystem, 1LPI, CH-3

IP/O/A/0310/7C ES Logic Subsystem, 1RB Isolation, CH-5

IP/0/A/0310/8C ES Logic Subsystem, 2RB Isolation, CH-6

IP/O/A/0310/15 RB Pneumatic Isolation Valve Check

IP/O/A/0310/17 PM of Quick Exhaust Valves on PR-2, -3, -4,

0 ITEM 6

Review the containment isolation initiation design and procedures and prepare

and implement all changes necessary to cause containment isolation of

all lines

whose isolation does not degrade core cooling capability upon automatic initia

tion of safety injection.

Response

All containment isolation valves with the exception of the following are normally

closed during operation and receive a close signal upon activation

of containment

isolation at a Reactor Building pressure of 4 psig:

(a) Pressurizer sample lines (1/2") are kept open to facilitate routine sampling.

(b) The low pressure service water to the reactor coolant pump motors

and component

cooling water to the reactor coolant pump seals are open. Closure of valves

in these systems could result in the reactor coolant pumps being

inoperable.

Such a condition is not considered generally desirable and is not consistent

with other portions of E Bulletin 79-O5A covering RCP operation for core

cooling capability.

No changes are deemed necessary to the containment isolation system design or

operation.

ITEM 7

For manual valves or manually-operated motor-driven valves which could defeat

or compromise the flow of auxiliary feedwater to the steam generators, pre

pare and implement procedures which:

(a) require that such valves be locked in their correct position, or

(b) require other similar positive position controls.

Response

The response to this item will be provided by April 16, 1979.

ITEM 8

Prepare and implement immediately procedures which assure that two independent

steam generator auxiliary feedwater flow paths, each with 100% flow capacity,

are operable at any time when heat removal from the primary system is

through the steam generators. When two independent 100% capacity flow paths

are not available, the capacity shall be restored within 72 hours or the

plant shall be placed in a cooling mode which does not rely on steam

generators for cooling within the next 12 hours.

When at least one 100% capacity flow path is not available, the reactor

shall be made subcritical within one hour and the facility placed in a

shutdown cooling mode which does not rely on steam generators for cooling

-rithin 12 hours or at the maximum safe shutdown rate.

Response

The response to this item will be provided by April 16, 1979.

ITEM 9

Review your operating modes and procedures for all systems designed to transfer

potentially radioactive gases and liquids out of the primary containment to

assure that undesired pumping of radioactive liquids and gases will not occur

inadvertently.

In particular, ensure that such an occurrence would not be caused by the resetting

of engineered safety features instrumentation. List all such systems and indicate:

a. Whether interlocks exist to prevent transfer when high radiation

indication exists, and

b. Whether such systems are isolated by the containment isolation signal.

Response

The following systems have the capability to transfer potentially radioactive

gases and liquids out of containment:

(a) Reactor Building normal sump to the miscellaneous waste holdup tank

(liquid) (b) Reactor Building emergency sump to the high activity waste tank (liquid)

(c) Reactor Building purge to the unit vent (gaseous)

(d) Gaseous waste disposal system (gaseous)

(e) Quench Tank drain system

The only system in which automatic actuation is a design feature is (a) above.

The pumps in this system have the capability to start automatically when sump

level is high and pump liquid to the miscellaneous holdup tank. However, operating

procedures require that this system be operated manually and a modification is in

process to remove this capability. The remaining systems are all operated manually,

with operator actuation required to align the system for transfer and then to

initiate transfer.

An interlock exists between the Reactor Building gas monitor

and Reactor Building

normal sump pump to prevent transfer in the event of high gaseous activity.

An interlock also exists between the unit vent gas monitor and Reactor Building

purge to secure the purge in the event of high gaseous activity in the unit vent.

All of these systems are isolated by the containment isolation signal. Upon re

setting of the containment isolation signal, all valves remain closed. Operator

action is required to reposition any containment isolation valve.

ITEM 10

Review and modify as necessary your maintenance and test procedures to ensure

that they require:

(a) Verification, by inspection, of the operability of redundant safety

related systems prior to the removal of any safety-related system from

service.

(b) Verification of the operability of all safety-related systems when they

are returned to service following maintenance or testing.

(c) A means of notifying involved reactor operating personnel whenever a

safety-related system is removed from and returned to service.

Response

The response to this item will be provided by April 16, 1979.

ITEM 11

All operating and maintenance personnel should be made aware of the extreme

seriousness and consequences of the simultaneous blocking of both auxiliary

feedwater trains at the Three Mile Island Unit 2 plant and other actions

taken during the early phases of the accident.

Response

The response to this item will be provided by April 16, 1979.