ML19208D061

Forwards Responses to IE Bulletin 79-08, Events Relevant to BWRs Identified During TMI Incident. Cold License Candidates Have Participated in TMI Review.Simulation Refresher Course Conducted

ML19208D061
Person / Time
Site:	Zimmer
Issue date:	08/07/1979
From:	Borgmann E CINCINNATI GAS & ELECTRIC CO.
To:	James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
NUDOCS 7909270803
Download: ML19208D061 (11)

Text

.

n t h jGLf &b k THE CINCINNATI GAS & ELECTRIC COMPANY rN #

CIN CIN N AT8,CMf C 4 52Cl August 7, 1979 c.A. soma u ww U.S. Nuclear Regulatory Cor=21ssion Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 ATTN: Mr. James G. Keppler, Director RE: Ef. H. ZIMMER NUCLEAR POWER STATION - UNIT 1 NRC IE BULLETIN 79-08 EVENTS RELEVANT TO BOILING WATER POWER REACTORS IDENTIFIED DURING THREE MILE ISLAND INCIDDIT W.O. 57300, JOB E-5590, FILE # 956, DOCKET # 50-358 Gentlemen:

The attached docu=ent is furnished in response to IE Bulletin 79-08.

We believe this infor=ation provides a complete response to NRC IE Bulletin 79-08.

Very truly yours, THE CINCINNATI GAS & ELECTRIC COMPAIN v-E.A. BORGMATI, SR. VICE PRESIDEIT HCB/kjd cc: U.S. Regulatory Co==1ssion Office of Inspection and Enforce =ent Division of Reactor Operations Inspection Washington, D.C. 20555 W.W. Schwiers S.G. Salay J.R. Schott W.D. Way=1re J.D. Flynn H.C. Brink = ann K.K. Chitkara W.E. Smith-San Jose R.J. Pruski General File AUP 131979 17909270[O]

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RESPONSES TO IE BULLETIN 79-08 ITEM l e Review the description of circu= stances described in Enclosure 1 of IE Bulletin 79-05 and the preliminary chronology of the TMI-2 3/28/79 accident included in Enclosure 1 to IE Bulletin 79-05A.

a. This review should be directed toward understanding:

(1) the extreme seriousness and consequences of the shnataneous blocking of both trains of a safety system at the Three Mile Island Unit 2 plant and other actions taken during the early phases of the accident; (2) the apparent operational errors which led to the eventual core damage; (3) the necessity to syste=atically analyze plant conditions and parameters and take appropriate corrective action.

b. Operational personnel should be instructed to (1) not over-ride autonatic action of engineered safety features unless continued operation of engineered safety features will re-sult in unsafe plant conditons (see Section 5a of this bulletin); and (2) not make operational decisions based solely on a single plant para =eter indication when one or more confirmatory indications are available.

c. All licensed operators and plant management and supervisors with operational responsibilities shall participate in this .

review and such participation shall be documented in plant .

records.

RESPONSE TO ITEM 1 All cold license candidates including operators, plant management and super-visors with operational responsibilities have participated in a review of the Three Mile Island Accident. This included the preliminary chronology of the TMI-2, 3/23/79 accident included in I&E Bulletin 79-05A, Enclosure I.

RESPONSE TO ITEM la The circumstances described in I&E Bulletin 79-05, Enclosure I and the understanding of subjects discussed in ISE Bulletin 79-08 Item Ia are being reviewed as follows:

a. A cold license candidate simulator refresher course was conducted in July and August, 1979. The course rein-forced and demonstrated Bk'R level instrumentation de-sign, interpretation, minor transients and upset con-ditions degrading to loss of coolant conditicas. Also covered operator decisions to preclude emergency system component operation.

b. A formal presentation of the events leading to and chronology as now known; with lessons learned be complete by Nov. 1, 1979 1042 157

. RESPONSE TO ITEM la CONT'D

c. The continuing onsite training program, Phase II, will provide additional review with operating licensed sup-ervisory and management personnel as further infomation is made available.

RESPONSE TO ITEM lb Station Administrative Directives (SAD's) have been revised to instruct operational personnel that automatic action of engineered safety features and isolation signals shall not be manually overridden unless:

a. Continued operation of the engineered safety features or isolation signals will result in unsafe plant con-ditons, or

b. It is known or positively determined that the automatic action was initiated by a spurious or erroneous signal and it is verified that operaton of the engineered safety feature or isolation is not required, or

c. Approved procedures specifically allow manual override under specific conditions, and those conditions are veri-fied to be satisfied.

Additionally, these points will be periodically restressed during the op-erator requalification training program.

SAD's have been revised to instruct operational personnel that when one or ,

more confirmatory indicators are available, operational decisions shall not be made based solely on a single plant parameter indication. Additionally, the SAD provides instructions to operational personnel that all available information should be considered in decisions to manually initiate, ters-inate, or controlioperation of safety systa=s.

RESPONSE TO ITE{ le Attendance during the review described in la, above was documented. wg ITEM 2 Review the containment isolation initiation design and procedures, and prepare and i=plement all changes necessary to initiate containment is-olation, whether manual or automatic, of all lines whose isolation does not degrade needed safety features or cooling capability, upon automatic initiation of safety injection.

RISPONSE TO ITEM 2 Containment isols. tion design and procedures have been reviewed. Containment isolation of all lines whose isolation does not degrade needed safety features or cooling capability is initiated either automatically cr manually upon auto-matic initiation of safety injection.

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ITEM 3 Describe the actions, both automatic and manual, necessary for proper functioning of the auxiliary heat removal systems (e.g., RCIC) that are ,

used when the main feedvater system is not operabic. For any =anual action necessary, describe in su==ary form the procedure, by which this action is taken in a ti=aly sense.

RISPONSE TO ITEM 3 The auxiliary heat removal systems provided to remove decay heat from the reactor core and contai= ment following loss of the feedvater system are:

High Pressure Core Spray (HPCS) System Reactor Core Isolation Cooling (RCIC) System Safety Relief Valves (SRV) and Automatic De-pressurination System (ADS)

Low Pressure Core Spray (LPCS) System Low Pressure Coolant Injection (LPCI) Mode of the Residual Heat Re= oval (RHR) System The operation of systa=s needed to achieve initial core cooling, contain=ent cooling, and extended core cooling for long term plant shutdown is decribed below,

a. INITIAL CORE COOLING Following loss of feedvater and subsequent reactor 2 scram, a low reactor water level signal vill auto-matica11y initiate main steam line isolation valve closure.

The safety relief valves (SRV's) vill automatically actuate to maintain reactor pressure. At the same ti=e, the lov vater level signal cutomatically initiates the HPCS and RCIC Syste=s. These systems vill continue to inject water into the reactor vessel until a high water level signal closes the HPCS injective valve and trips the RCIC system. Following a high reactor water level trip, the EPCS injection valve vill again reopen when reactor vacer level decreases to the low water level setpoint.

The RCIC System must be manually reset before it will reinitiate af ter a high water level trip. The EPCS and RCIC Systems have redundant supplies of water, normally taking suction from the cycled condensate storage tanks (CST's). The HPCS and RCIC Systems suction vill auto-matically transfer from the CST to the suppression pool if the CST vater is depleted or if the suppression pool water level increases to a high level.

The licensed operator can manually initiate the EPCS and RCIC Systems from the main control room before the low reactor water level automatic initiation level is reached.

The operator has the option of manual control af ter auto-matic initiation and can maintain reactor water level by throttling system flow rates. This would prevent a trip of the systems due to high water level. The operator can verify that these systems are delivering water to the re-actor vessel by any or all of the listed methods.

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ITQ4 3a CONT'D

1. Verifying reactor water level increases when systems initiate using recundant level indi-cators.

2. Verifying system flow rates using flow indicators in the control room.

3. Verifying system flow is to the reactor by checking control room position indication of motor-operated valves. This assures no diversion of system flow from the reactor.

The HPCS and RCIC Systems can maintain reactor wate level at full reactor pressure and until pressure decreases to where low pressure cystems such as the LPCI Mode of the RHR or Lov Pressure Core Spray (LPCS) can maintain reactor water level.

b. CONTAINMENT COOLING After reactor scram and isolation and establish =ent of sat-isfactory core cooling, the operator would initiate the suppression pool cooling mode of RHR. This mode of opera-tion removes heat resulting frcm safety relief valve (SRV) discharge and/or RCIC exhaust to the suppression pool. This '

is accomplished by placing one loop of the RHR System in

• the suppression pool cooling mode; (RER suction from and discharge to the suppression pool through one RER heat exchanger.)

The operator verifies proper operation of the RHR System containment cooling function from the main control room by:

1. Verifying RHR and Service Water (WS) Systam flow using system control room flow indicators.

2. Verifying correct RER and Service Water System flow paths using control room position indication of motor-operated valves.

3. Monitoring suppression pool water camperature.

Even though one loop of the RER is in the Suppression pool cooling mode, core cooling is its primary function. Thus, if a high dryvell pressure or low water level signal is re-ceived at any time during the period when the RER is in the suppression pool cooling mode, the RHR system till automat-ically revert to the LPCI injection mode. In addition, the HPCS and LPCS Systems would automatically start upon receipt of all ECCS initiation signals (s). The HPCS System functions as described in response 3.a and immediately injects water into the reactor vessel to maintain reactor water inventory.

The LPCS and LPCI Mode of RHR vould inject water into tie re-actor vessel if reactor pressure is below the respective 1042 160

ITEM 3b CONT'D system discharge pressures. Upon receipt of conincident low reactor water level and high drywell pressure signals, and after a two minute time delay, the Auto =atic Depressur-ization System (ADS) would relieve reactor pressure to allow the low pressure systems (LPCS & LPCI) to inject water into the vessel. (Also see responses 5.a and 5.b)

c. EXTENDED CORE COOLING When the reactor has been depressurized, the RHR System can be placed in the long ters shutdown cooling mode. The op-erator manually terminates the LPCI mode of one RER loop and places that loop in the shutdown cooling mode as follows:

1. Trip the selected RHR pump

11. Close motor operated valves (MOV's) in the suppression pool suction and discharge lines of the selected loop.

iii. Open the RHR shutdown cooling suction and dishcarge MOV's iv. Restart the selected RHR pu=p In this operating mode, the RHR System can coci the reactor to cold shutdown. Proper operation and flow paths in this mode can be verified by methods si=11at to those described for the containment cooling pode.

ITEM 4 Describe all uses and types of vessel level indication for both automatic and -

manual initiation of safery systems. Describe other redundant instrumentation which the operator might have to give the same information regarding plant status. Instruct operators to utilize other available information to initiate safety systems.

RESPONSE TO ITEM 4 Description of the reactor vessel level automatic initiation for the safety systens is previded in Chapter 7.3 of the FSAR. The description of the vessel level indication for manual initiation of the cafety systems is de-scribed in Chapter 7.5 of the FSAR. As described in response to item 1, SAD's have been revised to provide instructions to operating personnel to utill:e all available information.

ITEM S Review the action directed by the operating procedures and training instructions to ensure that:

a. Operators do not override automatic actions of engineered safety features, unless continued operation operation of engineered safety features will result in unsafe plant conditions (e.g. vessel integrity).

b. Operators are provided additional infor=ation and instructions to not rely upon vessel level indication alone for manual actions, but to also examine other plant parameter indi-cations in ev,aluating plant conditions.

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RESPONSE TO ITEM Sa Approved station operating and training procedures review accomplished by September 15, 1979 to ensure that appropriate instructions clearly specify that operating personnel do not override automatic actions of engineered safety features unless continued operation of these systems will result in unsafe plant conditions. However, it has been determined that several valid reasons exist for allowing an override of an automatic initiation signal or shutdown of a system af ter it has been automatically initialed (see response to Item 1, above) . For example,

a. If ar automatic initiation of the EPCS System and RCIC System occurs, the operator is permitted to shutdown the HFCS System if the RCIC System is capable of maintaining vessel level. This is allowed to prevent a trip of both systems due to high water level. As noted in response to Item 3 of this Bulletin, a trip of the RCIC System requires manual operator action to reset.

b. The procedures allow the operator to manually override auto =atic actuation of the ADS if it has been deter =ined that adequate water level is being =aintained by the dPCS Iystem. In this case, LPCI or LPCS is not required and, :herefore, ADS actuation can be interrupted. This override is permitted to allow a controlled cooldown and depre;surization of the reactor and prevents injection of suapression pool water into the reactor when it is not '

required. -

c. The proceduras allow transfer of part or all of the RHR System from the LPCI mode of operation to the Suppression Pool Cooling or Shutdown Cooling modes of operation when adequate reactor water level is maintained with part of the RHR System and/or other systems. This is permitted to insure that suppression pool water temperature and containment pressure 11=1ts are oaintained and provides controlled cooldown of the primary system.

RESPONSE TO ITEM Sb The SAD concerning s.ation operations has been revised as stated in the re-sponse to item 1 to assure operators consider all available infor=ation in decisions to take manual action. Operating procedures for specific events do describe expected parameter indications. Clarification, and where appropriate amplification will be made to specific operating procedures that describe parameter indications. Changes to procedures will be included in appropriate portions of operator training sessions. However, it should be recognized that events may occur such that vessel level indication might be the only immediately obvious parameter affected. We are reluctant to issue instructions which might be considered contrary to the direc*ive for operators to believe and respond conservatively to instru=ent indications unless the indications are proven to be incorrect.

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. . ITEM 7 CONT'D In particular, ensure that such an occurence would not be caused by the resetting of engineered safety featuras instrumentation. List all such systems and indicate:

a. Whether interlocks exist to prevent transfer when high radiation indication exists, and

b. Whether such systa=s are isolated by the containment isolation signal.

c. The basis on which continued operability of the above features is assured.

RESPONSE TO ITEM 7 All systems designed to transfer potentially radioactive gases and liquids from the primary containment are provided with automatic isolation valves.

Isolation signals are initiated by a variety of reactor, containment or system conditions. The trip setpoints for these automatic isolation sig-nals are listed in Technical Specification Table 3.3.2-2. valve groups that are operated by these trips are listed in Technical Specification Table 3.3.2-1. These valve groups are listed in Technical Specification Table 3.6.3-1.

The containment radiation monitoring system is not part of the containment isolation systen with the exception of the main steam line radiation moni-tors. The system provides information to the operator for the manual con- '

trol of the primary containment syste=s.

• ITEM 8 Review and modify as necessary your maintenance and test procedures to en-sure that the require:

a. Verification, by test or 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. Explicit notification of involved reactor operational

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

RESPONSE TO ITEM Ba. b and e The removal of equipment from service is controlled administrative 1y and/or with the use of the WR. The procedures which deal with equipment isolation specifically reference the responsible individuals to the applicable technical specification.

.. _ - ... 1042 164

RESPONCE TO ITE4 8a. b and e CONT'D Maintenance performed on safety-related equipment is controlled by the WR.

The (Work Request) SAD assigns the Shitt Supervisor the responsibility to identify technical specification requirements that pertain to any main-tenance activity. Any post work surveillance te3 ting that is required is also identified on the WR.

Written authorization to begin safety related corrective maintenance and any surveillance testing must be obtained from the Shif t Supervisor. Re-coving any equipment from service must be reviewed and authorized by the Shif t Supervisor. In addition, station operators are the only personnel who remove equipment from service and return it to service. Records of equipment tagged out and jumper and lifted leads are maintained by the operations group.

LTDi9 Review your prompt reporting procedures for NRC notification to assure that NRC is notified within one hour of the time the reactor is not in a con-trolled or expected condition of operation. Further, at that time an open continuous NRC. co==unication channel shall be established and =aintained with RESPONSE TO ITEM 9 Prompt reporting procedures for NRC notification vill be revised to es-tablish a continuous open line of comcunication with the NRC as rapidly [

as possible in the event the reactor is not in a controlled or expected condition of operation. It is our intent to work with the Region III Office of Inspection and Enforcement in the development of continuous coccunication channels.

ITEM 10 Review operating modes and procedures to deal with significant amounts of hydrogen gas that may be generated during a transient or other accident that would either remain inside the primary system or be released to the containment.

RESPONSE TO ITEM 10 In the eventhydrogen gas is generated by metal water reaction or radiolysis, the following methods are available to relieve the gas from the reactor vessel (primary system).

a. SAFETY RELIEF VALVES (SRV's)

There are thirteen SRV's located on the main steam lines that relieve to the quenchers located below the suppression pool water level. Since there is about 20 feet between the top of the core and the main steam line nozzles, a large volume of noncondensable gas can be relieved to the suppression pool via this pathway.

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e

. RESPONSE TO ITEM 10 CONT'D

b. LOSS OF C00 TANT A?CIb M

, A direct leakage path to the primary containment is created for release of noncondensables for certain postulated line rup ture s.-

c. REACTOR READ VENT The reactor vessel head vent relieves directly from the top of the vessel head via remote manual control from the main control room. The vent is directly piped to the reactor building equipment drain tank, and under supervision of a licensed operator or senior operator, the MOV's can be op-erated to relieve noncondensable gas to the primary contain-ment.

Af ter venting the hydrogen gas from the reactor vessel to the primary containment, the condensation of hysrogen and oxygen is continuously monitored by redundant trairs of contain=ent monitors. The Prinary Containment Combustable Gas Control System is activated remotely from the main control r::m, and this system, through the use of hydrogen recombiners, can adequately handle the postulated volume of hydrogen gas gen-erated from radiolysis and/or metal water reaction.

Operating procedures addressing the generation of hydrogen gas in the reactor vessel and release of hydrogen gas to the '

primary containment vill be reviewed by December 15, 1979. '

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