ML19269D614

Responds to IE Bulletin 79-08 Re Items Identified During TMI-2 Event.All Vessel Level Instrumentation Was Reviewed, Appropriate Procedures Are Being Revised & Containment Air Dilution Sys Installed

ML19269D614
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	04/27/1979
From:	Moody D VERMONT YANKEE NUCLEAR POWER CORP.
To:	Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
WVY-79-49, NUDOCS 7906050281
Download: ML19269D614 (8)

Text

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VERMONT Y AN KEE NUCLEAR POWER CORPORATION SEVENTY SEVEN GROVE STREET NUTLAND. VERMONT 05701 REPLYTO:

ENGINEERING OFFICE April 27, 1979 TURNPlKE RO AD WESTBORO, M ASS ACHUSETTS 01581 TELEPHON 7- 66 90 t United States Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, Pennsylvania 19406 Attention: Office of Inspection and Enforcement Boyce H. Grier, Director

References:

(a) License No. DPR-28 (Docket No. 50-271)

(b) USNRC Letter to VYNPC dated April 14, 1979; IE Bulletin 79-08

Subject:

Events Relevant To Boiling Water Reactors Identified During Three Mile Island Incident In accordance with your request, Reference (b), we are hereby providing the following information. The responses below address, in sequence, the questions raised in Reference (b).

Question #1

1. Review the description of circumstances described in Enclosure 1 of IE Bulletin 79-05 and the preliminary chronology of the TMI-2, 3/38/79 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 simultaneous 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; and (3) the necessity to systematically analyze plant conditions and parameters and take appropriate corrective action.

b. Operational personnel should be instructed to (1) not override ,

automatic action of engineered safety features unless continued operation of engineered safety features will result in unsafe plant conditions (see Section 5a of this bulletin); and (2) not make operational decisions based solely on a single plant parameter indication when one or more confirmatory indications are available.

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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.

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United States Nuclear Regulatory Commission Page 2 April 27, P;79

Response

A special lecture series on this subject for all licensed operators and plant management was prepared, presented and documented during the past ten days. Additionally, as more information on this incident becomes available, it will be factored into the ongoing operator training program.

Question #2 Review the containment isolation initiation design and procedures, and prepare and implement all changes necessary to initiate containment isolation, whether manual or automatic, of all lines vnose isolation does not degrade needed safety features or cooling capability, upon automatic initiation of safety injection.

Response

A review of the containment isolation initiation design and procedures was conducted to assure containment isolation will be achieved on all lines, whose isolation does not degrade needed safety features or cooling capability, upon automatic initiation of safety injection.

Automatic initiation of high pressure and low pressure emergency core cooling systems is achieved upon receipt of either a high drywell pressure or a ** low-low reactor vessel water level signal. All valves not essential to safety features or cooling capability isolate prior to, or simultaneous with, initiation of emergency cooling systms when such initiation is caused by receipt of a low-low reactor vessel water level signal. Should high drywell pressure initiate safety injection systems, the aforementioned valves isolate concurrently with the following exceptions:

• • • Group 1: Main Steam Line Isolation Valves including the Recirculation Loop Sample Line Valves and the Main Steam Line Drain Valves.

• • • Group 5: Reactor Cleanup System Isolation Valves.

The valves listed in Group 1 are closed upon any one of the following conditions:

1. Low-low reactor water level

2. High main steam line radiation 3 High main steam line flow

4. High main steam line tunnel temperature 22hl )l9

5. Low main steam line pressure (run mode only)

6. Condenser low vacuum

. . United States Nuclear Regulatory Commission Page 3 April 27, 1979 The receipt of a high drywell pressure isolation signal alone is not indicative of n rapid loss of vessel water inventory as compared to the low-low reactor water level condition. Unlike valves in Groups 2 through 4, the Group 1 valves remain open upon receipt of a high drywell pressure or a

• low reactor water level signal. This allows the removal of heat from the reactor core and assures that partial losses of feedwater supply would not unnecessarily initiate full isolation of the reactor, therby disrupting normal shutdown or recovery procedures and causing an unnecessary reactor transient. The design feature allows the removal of heat from the reactor following a scram without compromising isolation integrity should reactor vessel water level fall to the low-low setpoint. The additional five trip signals outlined above provide further assurance of isolation when such isolation is essential.

The Group 5 valves are closed upon receipt of a low reactor water level condition only. This setting, which is coincident with the low water level reactor scram setting, was selected to initiate isolation at the earliest indication of a possible breach in the nuclear system process barrier yet far enough below normal operating levels to avoid spurious isolation.

Valves which do not receive concurrent signals during automatic initiation of safety injection are provided with check valves to provide the necessary isolation capability as described in the Final Safety Analysis Report. A single exception is the Reactor Building Closed Cooling Water discharge from the primary containment. RBCCW is considered desirable to maintain containment cooling capabilities during an accident condition. This system does not communicate directly with the reactor coolant pressure boundary or containment free air space.

Manual isolation of this portion of the system can be accomplished from the station control room.

' Low reactor water level 127" above top of active fuel.

Low-low reactor water level 82.5" above top of active fuel.

• • • Primary Containment Valve Groups & Isolation Signals ar. described in Technical Specification pages 135-137 Question #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 feedwater system is not operable. For any manual action necessary, describe in summary form the procedure, by which this action is taken in a timely sense.

Response

A loss of feedwater transient has been analyzed and is delineated in Section 14.5 of the Final Safety Analysis Report. As stated, the reactor has been designed to scram automatically on low vessel water level; the main steam line isolation valves to close automatically, and the High Pressure Coolant Injection System (HPCI) and Reactor Core Isolation Cooling System (RCIC) to initiate automatically and prevent fuel damage.

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United States Nuclear Regulatory Commission Page 4 April 27, 1979 In addition, operating procedures are structured to prescribe manual initiation of these systems if sufficient operator response time exists before automatic action occurs. Although manual operator action was not essential, during at least one occasion (a loss of AC power with attendant loss of feedwater), station operators manually initiated HPCI in advance of automatic signals.

In summary, the operator actions necessary to manually initiate HPCI are: 1) to start the HPCI auxiliary oil and vacuum pumps; and, 2) to open the steam admission and water injection valves. RCIC can be manually initiated in a similar manner and either system can be operated through the entire range of reactor pressures until shutdown cooling can be established using the Residual Heat Removal system. Station operators also have other alternatives which consist of manually opening one or more main steam relief valves, thereby reducing reactor pressure and permitting makeup from the Condensate, Low Pressure Coolant Injection, or Core Spray Systems.

Question #4 Describe all uses and types of vessel level indication for both automatic and manual initiation of safety systems. Describe other redundant instrumentaton which the operator might have to give the same information rgarding plant status. Instruct operators to utilize other available information to initiate safety systems.

Response

In conjunction with the special review conducted per Item 1 of this bulletin, all types of vessel level indication were reviewed. This review of level indicators included types, uses, locations, and accident implications.

Additionally, vessel temperature indications were reviewed, examining types of instrument, location, and uses. Special emphasis was placed on reliability of instruments and backup sources of information if indications should fail (re. attached illustration entitled Vessel Level Indication).

Question #5 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 of engineered safety features will result in unsafe plant conditions (e.g. vessel integrity).

b. Operators are provided additional information and instructions to not rely upon vessel level indication alone for manual actions, but to also examine other plant parameter indications in evaluating plant conditions.

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United States Nuclear Regulatory Commission Page 5 April 27, 1979

Response

Station procedures provide operating personnel with information on symptoms and plant parameters useful in evaluating transient conditions and, provide precise instructions on the use of override and bypass controls. Additional instructions not to rely on a single vessel level indicator and not to override automatic actions of engineered safety features unless continued operation will result in unsafe plant conditions have been repeated within the past ten days.

Question #6 Review all safety-related valve positions, positioning requirements and positive controls to assure that valves remain positioned (open or closed) in a manner to ensure that proper operation of engineering safety features. Also review related procedures, such as those for maintenance, testing, plant and system startuo, and supervisory periodic (e.g.,

daily / shift checks,) surveillance to ensure that such valves are returned to their correct positions following necessary manipulations and are maintained in their proper positions during all operational modes.

Response

At Vermont Yankee sufficient management controls already exist to ensure proper valve position, performance, testing and system operability.

Historically, a supervisory review has always been required for each valve lineup, each component test, and each system operability test.

Additionally, the latter have been developed to prove total system performance, usually at rated flow conditions.

A very comprehensive watch relief by all shift personnel is also required including a walk-through of each control room panel, and a discussion of any equipment in an off-standard condition. Refinements to the valve lineup procedures are being considered to address even minor changes to valve positions which become necessary during routine operation and a formal procedure prescribing instrumentation rack valve lineups is being developed.

Question #7 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, venting or other release of radioactiv9 liquids and gases will not occur inadvertently.

In particular, ensure that such an occurrence would not be caused by the resett$ng of engineered safety features instrumentation. List all such syst ems and indicate:

a. Whether interlocks exist to prevent transfer when high radiation indication exists, and 2261 322

United States Nuclear Regulatory Commission Page 6 April 27, 1979

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

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

Response

a. There is no high radiation signal input into the isolation logic for any of the sytems designed to transfer potentially radioactive gases or liquids out of the primary containment. however, appropriate procedural precautions will be made to use existing radiation monitors to ensure that undesired pumping or venting of radioactive liquids and gases will not occur inadvertently as a result of resetting engineered safety features instrumentation.

b. The following transfer systems receive containment isolation signals:

1. Equipment and floor drain sumps -

Low reactor water level /hi drywell pressure.

2. Standby gas treatment system -

Low reactor water level /hi drywell pressure /hi-low radiation in reactor building ventilation or refueling floor.

3 RHR discharge to radwaste -

Low reactor water level /hi drywell pressure.

4. Containment atomsphere dilution system -

Low reactor water level /hi drywell pressure hi/ low radiation in reactor building ventilation or refueling floor.

c. Contair. ment isolation logic is tested at 6 month intervals and the sensors that input to the logic are tested monthly.

Question #8 Review and modify as necessary your maintenance and test procedures to ensure that they 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. Explict notification of involved reactor operational personnel whenever a safety-related system is removed from and returned to service.

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United States Nuclear Regulatory Commission Page 7 April 27, 1979

Response

We have reviewed our maintenance and test procedures and find that the issues of prior operability testing, post operability testing, and explict notification of operational personnel have been adequately addressed in A.P.0025, Plant Equipment Control; A.P.0153, Maintenance of Operations Departmental Logs, and A.P. 0140, Local Control Switching Rules. Nevertheless, because of the extreme seriousness and consequences of the TMI incident we have intensified this area through the following re finements:

1. The need for explict notification of control room personnel for all technical specification and safety related inspections, maintenance, and surveillance will be reemphasized to all station employees at a general staff meeting within the near future.

2. The shift supervisor's log will be restructured and broadened to identify on a continuing basis all equipment which is removed from service, when thie equipment is returned to service, and to provide a more comprehensive record of daily station activities.

3 A.P.0025, Plant Equipment Control, will be expanded to include technical specification and safety related equipment. Presently only the former is recognized. Appropriate departmental procedure changes are also being generated to include reference to, or incorporating the interface requirement of, A.P.0025.

Question #9 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 controlled or expected condition of operation. Further, at that time an open continuous communication channel shall be established and maintained with the NRC.

Response

The appropriate plant procedure (s) were reviewed and are in the process of being revised to assure Nuclear Regulatory Commission notification within one hour of the time the reactor is not in a controlled or expected condition of operation. The above procedure (s) will detail the requirement of establishing and maintaining an open continuous com=unication channel with the Commission.

Question #10 Review operating modes and procedures to deal with significant amounts of hydrogren 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.

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4 United States Nuclear Regulatory Commission Page 8 April 27, 1979

Response

A review of operating modes and procedures to deal with hydrogen gas generated during an accident has been initiated. Vermont Yankee does have installed capability to control the buildup of hydrogen in both the reactor vessel and the primary containment system.

Hydrogen gas in the reactor vessel can be controlled by venting to the primary contianment through either the safety / relief valves (SRV's) or the head vent, both of which can be manually activated from the control room. If an accident leads to significant metal-water reaction, the adGitional quantity of hydrogen would not be of particular concern in a BRA because it could not, by itself, lead to interference with post-incident core cooling. The venting paths mentioned above can be remotely activated from the station control room. The SRV's will also automatically open a high pressure or due to an Automatic Depressurization System (part of ECCS) activation signal.

Vermont Yankee has installed a Containment Air Dilution System to prevent the occurrence of combustible mixtures of hydrogen and air within the primary containment following an accident. Proposed Technical Specifications to provide assurance that the CAD System will be operable in the event it is required were submitted to the NRC on July 15, 1976.

To date, no reply has been received to that submittal. The Vermont Yankee staff, however, maintains surveillance on the CAD System in accordance with those proposed Technican Specifications and considers the system available if needed.

We trust that this information is satisfactory; however, should you have any further questions, please contact R. J. Wanczyk of this office.

Very truly yours, VERMONT YANKEE NUCLEAR POWER CORPORATION M

. E. Moody. W Manager of Operations

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