ML18037A311

Forwards Documentation of TMI Action Items Requiring 810101 Submittal,Per NRC 801031 Request.Sixteen Oversize Drawings Encl

ML18037A311
Person / Time
Site:	Nine Mile Point
Issue date:	12/31/1980
From:	Dise D NIAGARA MOHAWK POWER CORP.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
TASK-1.A.1.1, TASK-1.C.5, TASK-2.K.3.27, TASK-TM NUDOCS 8101060400
Download: ML18037A311 (170)

Text

REGULATE INFORMATION DISTRIBUTION STEM (RIDS)

ACCESSION NBR 8101060000 DOC ~ DATE 80/12/31 NOTARIZED NO DOCKET FACIL'.50 220 Nine Mile Point Nuclear Station< Unit iP Niagara Powe 050002?0 AUTH INANE AUTHOR AFFILIATION DISEUR D ~ P ~

Niagara Mohawk Power Corp.

REC IP, NAME RECIPIENT AFFILIATION EISENHUTED ~ G ~

Division of Licensing

SUBJECT:

forwards documentation of TMI action items requiring 810101 submittaliper 801031 requests Sixteen oversize drawings encl'ISTRIBUTION CODE:

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UMQHA~K NIAGARAMOHAWKPOWER CORPORATION/300 ERIE BOULEVARDWEST, SYRACUSE, N.Y. 13202/TELEPHONE (315) 474-1511 December 31, 1980 Mr. Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.

C.

20555

Dear Mr. Eisenhut:

Re:

Nine Mile Point Unit 1

Docket No. 50-220 DPR-63 Attached is the requested documentation for those TMI items required to be submitted by January 1,

1981 in your October 31, 1980 letter.

Also included is information for certain items which are planned to be implemented at Nine Mile Point Unit 1 before your schedule requires them.

Very truly yours, NIAGARA MOHAWK POWER CORPORATION PEF:ja Attachment 3>lh83q g(~ ',

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TMI ACTION PLAN ITEMS REQUIRING A JANUARY 1, 1981 SUBMITTAL NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT UNIT 1

DOCKET NO. 50-220 DPR-63 KNLNM)'00NHFll.f Mm Control fP S]olci6~ooo

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TABLE OF CONTENTS ITEM NO.

I.A.1.1 I.c. 1 I.C.5 I.C.6 I I.',B.. 2 II.!B '.4 II.E.4.2 I I.F.1.2 II.F.1.3 II.F.1.4 II.F. 1. 5 II.K.3.3 I I.K.3. 13 I I.K.3.14 I I.K.3.15 I I.K.3.17 II.K.3.19 II.K.3.21 II.K.3.22 II.K.3.24 II.K.3.25 II.K.3.27 TITLE Shift Technical Advisor Short Term Accident and Procedures Review Feedback of Operating Experience Verify Correct Perf ormance of Operating Activities Plant Shielding Training for Mitigating Core Damage Containment Isolation Dependability Iodine/Particulate Sampling Containment High Range Monitor Containment Pressure Containment Water Level Reporting Safety Valve and Relief Valve F a ilures and. Chall eng es HPCI and RCIC Initiation Levels Isolation Condenser Isolation Isolation of HPCI and RCIC ECC System Outages Interlock on Recirculation Pump Loops Restart of Core Spray and LPC I RCIC Suction Space Cooling for HPCI and RCIC Power on Pump Seals Cordon Reference Level PAGE 10 13 16 17 18 19 20 21 22 23 24 25 26

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TABLE OF CONTENTS (Continued)

ITEM NO.

TITLE PAGE II.K.3.28 II.K.3.29 I I.K.3.44 Qualification of ADS Accumulators 27 Performance of Isolation Condensers 28 Evaluation of Transients with Sing 1 e 29 Failure I I.K.3.45 II.K.3.46 III.A.2 III.D.3.3 III.D.3.4 Manual Depressuri zation Michelson Concerns Emergency Preparedness Inplant Radiation Monitoring Control Room Habitability 30 31 32 33 34

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TMI ACTION PLAN ITEM NO. I.A.1.1 SHIFT TECHNICAL ADVISOR NRC POSITION Each licensee shall provide an on-shift technical advisor to the shift supervisor.

The shift technical advisor (STA) may serve more than one unit at a multiunit site if qualified to perform the advisor function for the various units.

The STA shall have a bachelor's degree or equivalent in a scientific or engineering discipline and have received specific training in the response and analysis of the plant for transients and accidents.

The STA shall also receive training in plant design and layout, including the capabilities of instrumentation and controls in the control room.

The licensee shall assign normal duties to the STAs that pertain to the engineering aspects of assuring safe operations of the plant, including the review and evaluation of operating experience.

RESPONSE

Since January 7,

1980, an Assistant Shift Supervisor (Shift Technical Advisor) has been added to the normal Control Room shift composition to be a non-shift technical advisor to the shift supervisor.

The Assistant Shift Supervisors have a bachelor's degree or equivalent in a scientific or engineering discipline.

The operations experience assessment function is performed by special meetings of the Site Operations Review Committee which are held at least once every two months.

These meetings are attended by the Assistant Station Shift Supervisor as available.

Training which meets the lessons learned requirements has been completed (i.e.

training in the response and analysis of the plant for transients and accidents and in plant design and layout, including the capabilities of instrumentation and controls in the control room).

Attachment 1 included at the end of this report provides a description of the current training program for the Nine Mile Point Unit 1 Shift Technical Advisors and the long term training requirements.

Also included is a

comparison of our training program to the draft INPO document entitled "Nuclear Power Plant Shift Technical Advisor - Recommendations for Position Description, gualifications, Education and Training."

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TMI ACTION PLAN ITEM NO. I.C.1 GUIDANCE FOR THE EVALUATION AND DEVELOPMENT OF PRO EDURE FOR. TRAN IENT AND AC IDENT NRC '

POSITION In letters of September 13 and 27, October 10 and 30, and November 9,

1979, NRR required licensees of operating plants, operating license applicants, and licensees of plants under construction, to perform analyses of transients and accidents, prepare emergency procedure guidelines, upgrade emergency procedures, including procedures for operating with natural circulation conditions, and to conduct operator retraining (see also Item I.A.2.1).

Emergency procedures are required to be consistent with the actions necessary to cope with the transients and accidents analyzed.

Analyses of transients and accidents were to be completed in early 1980 and implementation of procedures and retraining were to be completed three months after emergency procedure guidelines were established;

however, some difficulty in completing these requirements has been experienced.

Clarification of the scope of the task and appropriate schedule revisions are being developed.

In the course of review of these matters on BEW designed

plants, the staff will follow up on the Bulletin and Orders matters relating to analysis methods and results, as listed in NUREG-0660 Appendix C.

See Table C.l, Items 3, 4, 16, 18, 24, 25, 26, 27; Table C.2, Items 4, 12, 17, 18, 19, 20; and Table C.3, Items 6, 35, 37, 38, 39, 41, 42, 47, 55, 57.

RESPONSE

Niagara Mohawk has met the requirements of Item I.C.1 to perform revised analyses and prepare emergency procedure guidelines by January 1,

1981 through the BWR Owners Group by submittal of the following documents:

1.

"Additional Information Required for NRC Staff Generic Report On Boiling Water Reactors",

NED0-24708, August 1979.

2.

Section 3.2.1 (Revised) of NED0-24708, "Analysis of Loss of Feedwater Events" transmitted by R.H. Buchholz's letter to D.F. Ross dated March 31, 1980.

3.

BWR Emergency Procedure Guidelines Revision 0 (Prepublication form),

transmitted by R.H. Buccholz's letter to D.G. Eisenhut dated June 30, 1980.

4.

Section 3.2.2 of NED0-24708, "Other Operational Transients" transmitted by R.H. Buccholz's letter to D.G. Eisenhut dated August 22, 1980.

5.

Section 3.5.2.1 (Revised) of NED0-24708, "Analysis to Demonstrate Adequate Core Cooling" and Section 3.5.2.4 of NED0-24708, "Justification of Analysis Methods", transmitted by R.H. Buccholz's letter to D.G. Eisenhut dated September 16, 1980.

Upgrading of emergency procedures and operator training will be completed as required following the Nuclear Regulatory Commission's staff review of the BWR Emergency Procedure Guidelines.

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TMI ACTION PLAN ITEM NO. I.C.5 PROCEOURES FOR FEEDBACK OF OPERATING EXPERIENCE TO PLANT STAFF NRC POSITION In accordance with Task Action Plan I.C.5, Procedures for Feedback of Operating Experience to Plant Staff (NUREG-0660),

each applicant for an operating license shall prepare procedures to assure that operating information pertinent to plant safety originating both within and outside the utility organization is continually supplied to operators and other personnel and is incorporated into training and retraining programs.

These procedures shal 1:

(1) Clearly identify organizational responsibilities for review of operating experience, the feedback of pertinent information to operators and other personnel, and the incorporation of such information into training and retraining programs; (2) Identify the administrative and technical review steps necessary in translating recommendations by the operating experience assessment group into plant actions (e.g.,

changes to procedures; operating orders);

(3) Identify the recipients of various categories of information from operating experience (i.e., supervisory personnel, shift technical

advisors, operators, maintenance personnel, health physics technicians) or otherwise provide means through which such information can be readily related to the job functions of the recipients; (4) Provide means to assure that affected personnel become aware of and understand information of sufficient importance that should not wait for emphasis through routine training and retraining programs; (5) Assure that plant personnel do not routinely receive extraneous and unimportant information on operating experience in such volume that it would obscure priority information or otherwise detract from overall job performance and proficiency; (6) Provide suitable checks to assure that conflicting or contradictory information is not conveyed to operators and other personnel until resolution is reached;

and, (7) Provide periodic internal audit to assure that the feedback program functions effectively at all levels.

~Res ense A procedure has been issued and is available for review by the Nuclear Regulatory Commission at the Nine Mile Point Unit 1 site.

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TMI ACTION PLAN ITEM NO. I.C.6 GUIDANCE ON PROCEDURES FOR VERIFYING CORRECT PERFORMANCE OF OPERATING ACTIVITIES NRC POSITION It is required (from NUREG-0660) that licensees'rocedures be reviewed and

revised, as necessary, to assure that an effective system of verifying the correct performance of operating activities is provided as a means of reducing human errors and improving the quality of normal operations.

This will reduce the frequency of occurrence of situations that could result in or contribute to accidents.

Such a verification system may include automatic system status monitoring, human verification of operations and maintenance activities independent of the people performing the activity (see NUREG-0585, Recoranendation 5), or both.

Implementation of automatic status monitoring if required wi 11 reduce the extent of human verification of operations and maintenance activities but will not eliminate the need for such verification in all instances.

The procedures adopted by the licensees may consist of two phases--one before and one after installation of automatic status monitoring equipment, if required, in accordance with item I.D.3.

Response

Procedures have been revised and are available for review by the Nuclear Regulatory Commission at the Nine Mile Point Unit 1 site.

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TMI ACTION PLAN ITEM NO. II.B.2 DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION N

WHICH MAY BE USED IN POST ACCIDENT OPERATIONS NRC POSITION With the assumption of a postaccident release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalent of 50K of the core radioiodine, 100/ of the core noble gas inventory, and lX of the core solids are contained in the primary coolant),

each licensee shall perform a radiation and shielding-design review of the spaces around systems that may, as a result of an accident, contain highly radioactive materials.

The design review should identify the location of vital areas and equipment, such as the control room, radwaste control stations, emergency power supplies, motor control centers, and instrument

areas, in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during postaccident operations of these systems.

Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design

changes, increases permanent or temporary shielding, or postaccident procedural controls.

The design review shall determine which types of corrective actions are needed for vital areas throughout the facility.

RESPONSE

Niagara Mohawk's submittals of December 31,

1979, January 31,

1980, June 20, 1980 and September 17, 1980 provided the results of the plant shielding design review and proposed modifications for Nine Mile Point Unit l.

The shielding design review for Nine Mile Point Unit 1 did not include analyses of LOCA events in which the primary system remains pressurized.

This is because the plant design and the emergency procedures would lead to depressurization of the system and injection of low pressure cooling water before fuel failures and resulting fission product releases would occur.

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TMI ACTION PLAN ITEM NO. II.B.4 TRAINING FOR MITIGATING CORE DAMAGE NRC POSITION Licensees are required to develop a training program to teach the use of installed equipment and systems to control or mitigate accidents in which the core is severely damaged.

They must then implement the training program.

RESPONSE

As indicated in our letter of December 17, 1980, the training program will not be available for submittal to the Nuclear Regulatory staff until April 1, 1981.

Due to the spring 1981 refueling outage implementation of the training program will not begin until after the outage.

The training will be completed by December 31, 1981.

The shift technical advisors and all operating personnel who held senior reactor operator or reactor operator licenses from the plant manager through the operations chain to the shift operators will participate in this training program.

Managers and technicians in the Instrumentation and Control

( IEC),

health physics and chemistry departments will receive training commensurate with their responsibilities.

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TMI ACTION PLAN ITEM NO. II.E.4.2 CONTAINMENT ISOLATION DEPENDABAILITY NRC POSITION (1) Containment isolation system designs shall comply with the recommendations of Standard Review Plan Section 6.2.4 (i.e., that there be diversity in the parameters senses for the initiation of containment isolation).

(2) All plant personnel shall give careful consideration to the definition of essential and nonessential

systems, identify each system determined to be essential, identify each system determined to be nonessential, describe the basis for selection of each essential

system, modify their containment isolation designs accordingly, and report the results of the reevaluation to the NRC.

(3) All nonessential systems shall be automatically isolated by the containment isolation signal.

(4) The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves.

Reopening of contain-ment isolation valves shall require deliberate operator action.

(5) The containment setpoint pressure that initiates containment isolation for nonessential penetrations must be reduced to the minimum compatible with normal operating conditions.

(6) Containment purge valves that do not satisfy the operability criteria set forth in Branch Technical Position CSB 6-4 or the Staff Interim Position of October 23, 1979 must be sealed closed as defined in SRP 6.2.4, item II.3.f during operational conditions 1, 2, 3, and 4.

Furthermore, these valves must be verified to be closed at least every 31'days.

(A copy of the Staff Interim Position is enclosed as Attachment 1).

(7) Containment purge and vent isolation valves must close on a high radiation signal.

RESPONSE

Items 1, 2, 3 and 4 of the Nuclear Regulatory Commission's POSITION above were addressed in Niagara Mohawk's December 31, 1979 submittal wKicMdocumented our review of the containment isolation provisions of Nine Mile Point Unit 1.

Deviations from the NRC's documented position were identified and justified.

Based upon this justification, no modifications were proposed for the Nine Mile Point Unit 1.

The deviations and the justification are identified below:

1.

The Main Steam (including warm-up and emergency cooling vents),

Reactor Cleanup and Shutdown Cooling lines isolate on the low low reactor vessel water level containment isolation signal.

Isolation of these systems is not initiated on high drywell pressure because they are closed systems capable of handling radioactivity levels associated with normal operation.

Abnormally high levels of radioactivity could result from fuel damage caused by the reactor water level dropping below the top of the fuel.

Isolation of these systems would occur at low low water level setpoint which is approximately 7 feet 6 3/4 inches above the top of fuel.

Therefore, isolation would occur before any fuel failures.

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The Drywell and Suppression Chamber Ng Make up and H2-02 sampling and Containment Airborne Activity Monitor Systems isolate on a containment isolation signal (i.e.

low low water level or high drywell pressure).

These systems are provided with overrides so that they can be manually reopened for controlled venting and purging and monitoring purposes.

The Reactor Building Closed Loop Cooling to the recirculation pump coolers and drywell coolers are non-essential systems which do not automatically isolate on containment isolation signals.

These are closed systems inside the drywell and are not connected to the reactor coolant pressure boundary or open to the free space of containment.

They provide cooling to the'on-safety related pump motors and drywell coolers which, although not required to mitigate the consequences of an accident, are beneficial if they continued to operate.

The supply lines are provided with a self-actuating check valve and the return lines are provided with a blocking valve which can be remotely closed from the control room.

In addition, the supply line to the recirculation pump coolers has a blocking valve inside containment which can be remotely operated in the Reactor Building.

Therefore, these systems can be isolated if high radiation leakage into the systems occurs.

Remote manual isolation of the return line is acceptable based on General Design Criterion 57-Closed S

stem Isolation Valves which states that "Each line that penetrates primary reactor containment and is neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere shall have at least one containment isolation valve which shall be either automatic or locked closed or capable of remote manual operation."

The use of check valves as the automatic isolation valves outside containment as in the supply lines discussed above has been previously justified in Niagara Nohawk's Technical Supplement to Petition for Conversion from Provisional Operatsno License o

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The Atmosphere to Suppression Chamber Vacuum Relief line contains an air'per'ated/DC solenoid valve and a self-activating check valve.

The air operated/DC solenoid valve is a normally closed valve which will open on a negative pressure relative to atmosphere.

The air operated/DC solenoid valve does not receive the automatic containment isolation signals.

This is considered acceptable since this valve would not be normally opened and the self-activating check valve will prevent flow from the torus to atmosphere.

The Suppression Chamber Water Makeup line has a diaphragm operated DC solenoid valve and a self-actuating check valve.

The diaphragm operated DC solenoid valve is a normally closed valve which is remotely operable from the Control Room.

Although not identified as a deviation from the NRC's position in our December 31, 1979 submittal, this isolation valve does not receive an automatic containment isolation signal.

This is considered acceptable since this valve would not be normally opened but may be required to be opened to provide make up to the torus during an accident in which the containment is isolated.

The self-actuating check valve will prevent flow from the torus out the make-up line.

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In addition to the above a review of the isolation of lines penetrating the primary containment at Nine Mile Point Unit 1 has been performed in accordance with the requirements of General Design Criteria (GDC) 55, 56 and 57.

Table 1

attached at the end of this report is a penetration by penetration listing of all lines penetrating containment.

The last column in the table documents the compliance with GDC 55, 56 or 57.

As a result of this review, the following modifications/changes have been recommended.

1.

The service water and breathing air connections for the drywell (penetrations X-122 and 121 respectively) will have the inside manual valve changed to a normally locked closed valve.

This will be performed prior to start up from the spring 1981 refueling outage.

2.

The recirculation system sample line and the containment spray test line to waste disposal will be provided with automatic isolation valves.

As indicated in our response to I. E.Bulletin 79-08, these modifications will be performed during the spring 1981 refueling outage.

The existing containment isolation pressure setpoint of 3.5 psig wi 11 continue to be used for initiating'containment isolation.

This is 1.3 psi higher than the'aximum observed pressure inside containment over the past year during normal operation.

The existing containment isolation pressure setpoint is not significantly different from the setpoint recommended by Nuclear Regulatory Commission to warrant a Technical Specification change.

Therefore, Technical Specification changes reflecting a change in containment isolation pressure setpoint will not be submitted.

As indicated in our letter of December 17, 1979, the containment vent and purge line's at Nine Mile Point Unit 1 meets the Staff 's Interim Position of October 23, 1979 by limiting the outboard isolation valves -to 50 degrees maximum opening.

Therefore, no further action is required on this item at this time.

Isolation of the containment purge and vent valves on high radiation will be provided by January 1,

1982.

The isolation signal will be provided by the containment high range radiation monitors which are to be installed by January 1,

1982.

Technical Specification changes reflecting this addition will be submitted by June 30, 1981 for Nuclear Regulatory Commission staff approval.

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TMI ACTION PLAN ITEM NO. II.F.1.2 SAMPLING AND ANALYSIS OF PLANT EFFLUENTS NRC POSITION Because iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radioiodines for the accident condition shall be provided with sampling conducted by adsorption on charcoal or other media, followed by onsite laboratory analysis.

RESPONSE

Niagara Mohawk's submittal of December 17, 1980 indicated that procedures for the removal and analysis of samples would be reviewed and revised by January 1,

1981.

Although procedures have been drafted, the procedures will not be approved and issued until January 31, 1981.

The sampling of radioiodine and particulate is performed by removing the charcoal canister (for iodine) and particulate filter located in the sample line to the stack monitor followed by analysis in the lab.

This method of sampling is considered to be a continuous sampling method as a new charcoal canister and particulate filter will be installed when the others are removed for analysis.

By January 1,

1982 the sample holder for the charcoal canister and particulate filter will be modified so that highly radioactive samples can be removed.

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TMI ACTION ITEM NO. II.F.1.3 CONTAINMENT HIGH-RANGE RADIATION MONITOR NRC POSITION In containment radiation-level monitors with a maximum range of 108 rad/hr shall be installed.

A minimum of two such monitors that are physically separated shall be provided.

Monitors shall be developed and qualified to function in an accident environment.

RESPONSE

Niagara Mohawk plans to install two independent containment high-range radiation monitors during the spring 1981 refueling outage, but no later than January 1,

1982.

The monitors will be installed in existing spare penetration sleeves which will extend into the free space of the containment such that the entire active portion of the detector will be inside the containment.

The penetrations maximum thickness is approximately 1/4 inch.

Locating the monitors in penetration sleeves will increase the reliability of the monitors making them accessible for replacement, maintenance and calibration, while providing assessment of area radiation conditions inside containment.

The purpose of these monitors is to detect gross fuel failure.

The proposed in-sleeve arrangement will perform this function since early indicators of fuel damage have high energy gammas (such as Xe-138 (2 MEV), Kr-87 (2.5 MEV),

Kr-88 (2.4 MEV)).

Although NUREG 0737 indicates that these monitors should respond to energies as low as 0.060 MEV, low energy isotopes (such as Xe-133) are not significant as outlined below.

1.

Higher energy isotopes will dominate the fission product mix until they have decayed significantly.

Low energy isotopes will not be a

major constituent in the fission product mix until this occurs (i.e.,

several days after an accident).

2.

Since low energy isotopes will not require monitoring until several days after an accident drywell sampling can be utilized.

3.

Low energy isotopes such as Xe-133 are of little biological consequence because of their low energies and low gamma abundance.

The monitors to be installed at Nine Mile Pont Unit 1 are Model No. RD-23 Gamma Detector and RP-2C high-range radiation readout module signal processor supplied by the General Atomic Company.

The gamma detector has a range of 10o to 108 R/hr. It has been environmentally qualified to withstand

350oF, 70 psig and OX to 100K humidity and seismically qualified to IEEE 344-1975.

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TMI ACTION PLAN ITEM NO. II.K.3.15 MODIFY BREAK DETECTION LOGIC TO PREVENT SPURIOUS ISOLATION OF'IGH PRESSURE OOLANT INJE TI N AND R

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CORE I OLATION COOLING NRC POSITION The high-pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) systems use differential pressure sensors on elbow taps in the steam lines to their turbine drives to detect and isolate pipe breaks in the systems.

The pipe-break-detection circuitry has resulted in spurious isolation of the HPCI and RCIC systems due to the pressure spike which accompanies startup of the systems.

The pipe-break-detection circuitry should be modified so that pressure spikes resulting from HPCI and RCIC system initiation will not cause inadvertent system isolation.

RESPONSE

As indicated in our letters of June 20, 1980 and December 17, 1980, this item is not directly applicable to Nine Mile Point Unit 1 because it pertains to boiling water reactors with steam driven turbines in RCIC and HPCI systems.

The Nine Mile Point Unit 1 design does not have RCIC and HPCI systems with steam driven turbines.

Therefore, this item requires no further action for Nine Mile Point Unit l.

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TMI ACTION PLAN ITEM NO. II.K.3.17 REPORT ON OUTAGES OF EMERGENCY CORE - COOLING SYSTEMS LICENSEE RE ORT AND PROPOSED TECHNICAL PECIF ICAT ION CHANGES NRC POSITION Several Components of the emergency core-cooling (ECC) systems are permitted by technical specifications to have substantial outage times (e.g.,

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for one diesel-generator; 14 days for the HPCI system).

In addition, there are no cumulative outage time limitations for ECC systems.

Licensees should submit a report detailing outage dates and lengths of outages for all ECC systems for the last 5 years of operation.

The report should also include the causes of the outages (i.e., controller failure, spurious isolation).

~Res ense This requirement was met by our letter dated October 8, 1980 from Mr. D. P.

Disc to Mr. D.

G. Eisenhut.

This letter transmitted a report with the information requested above.

Included in the report were outage dates and duration of the outage, cause of the outage, system and component involved and the corrective action taken.

Niagara Mohawk did not propose any changes to improve the availability of emergency core cooling equipment since the number of outages was not significant enough to warrant changes.

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TMI ACTION PLAN ITEM NO. II.K.3.19 INTERLOCK ON RECIRCULATION PUMP LOOPS NRC POSITION Interlocks should be installed on nonjet pump plants (other than Humboldt Bay) to assure that at least two recirculation loops are open for recirculation flow for modes other than cold shutdown.

This is to assure that the level measurements in the downcomer region are representative of the level in the core region.

RESPONSE

As indicated in our letters of June 20, 1980 and December 17, 1980, Niagara Mohawk currently has administrative controls and Technical Specification requirements at Nine Mile Point Unit 1 to assure that at least two recirculation loops are open for recirculation flow for all operating modes other than cold shutdown.

Therefore, no further action on this item is required at Nine Mile Point Unit 1.

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TMI ACTION PLAN ITEM NO. II.K.3.21 RESTART OF CORE SPRAY AND LOW-PRESSURE COOLANT INJECTION SYSTEMS NRC POSITION The core spray and low pressure coolant injection (LPCI) system flow may be stopped by the operator.

These systems will not restart automatically on loss of water level if an initiation signal is still present.

The core spray and LPCI system logic should be modified so that these systems will restart, if required, to assure adequate core cooling.

Because this design modification affects several core cooling modes under accident conditions, a preliminary design should be submitted for staff review and approval prior to making the actual modification.

RESPONSE

The Nine Mile Point Unit 1 design includes two Low Pressure Core Spray

systens, but not a separate Low Pressure Coolant Injection System.

As indicated in our letters of June 20, 1980 and December 17, 1980, the core spray pumps will automatically restart following a manual stop upon receipt of a low-low water level or high drywell pressure signal (LOCA) or if one. or both of the signals is still present.

Although it is possible to place the core spray pump switches in the locked out mode, Niagara Mohawk does not believe that modification of the core spray system logic is required at Nine Mile Point Unit l.

This position is also set forth in the study performed by General Electric for the BWR Owner's Group.

This report titled NUREG 0737 Item II.K.3.21 Core S ray and Low Pressure Coolant Inspection Systems Leve Instiation was transmstte to t e NR y a etter ate Decem er rom Mr. D. B.

Waters, Chairman TMI BWR Owners Group to Mr. D.

G. Eisenhut.

Although this report is generic, the discussions regarding non-jet pump plants and the low pressure core spray system are applicable to Nine Mile Point Unit l.

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TMI ACTION PLAN ITEM NO. II.K.3.22 AUTOMATIC SWITCHOVER OF REACTOR CORE ISOLATION COOLING SYSTEM SUCTION - VERIFY PROCEDURES AND MODIFY DESIGN NRC POSITION The reactor core isolation cooling (RCIC) system takes suction from the condensate storage tank with manual switchover to the suppression pool when the condensate storage tank level is low.

This switchover should be made automatically.

Until the automatic switchover is implemented, licenses should verify that clear and cogent procedures exist for the manual switchover of the RCIC system suction from the condensate storage tank to the suppression pool.

RESPONSE

As indicated in Niagara Mohawk's letters of June 20, 1980 and December 17, 1980, this item is not applicable to Nine Mile Point Unit 1.

Nine Mile Point Unit 1 has a gravity fed closed loop Emergency Condenser system instead of a reactor core isolation cooling system.

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TMI ACTION PLAN ITEM NO. II.K.3.24 CONFIRM ADEQUACY OF SPACE COOLING FOR HIGH-PRESSURE COOLANT INJECTION AND REACTOR CORE ISOLATION COOLING SYSTEMS NRC POSITION Long-term operation of the reactor core isolation cooling (RCIC) and high-pressure coolant injection '(HPCI) systems may require space cooling to maintain the pump-room temperatures within allowable limits.

Licensees should verify the acceptability of the consequences of a complete loss of alternating-current power.

The RCIC and HPCI systems should be designed to withstand a complete loss of offsite alternating-current power to their support

systems, including coolers, for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

RESPONSE

As indicated in Niagara Mohawk's letter of June 20, 1980 and December 17,

1980, no action on this item is required for Nine Mile Point Unit 1.

The Nine Mile Point Unit 1 design does not include HPCI and RCIC systems with pump rooms which require space cooling to maintain temperatures within allowable limits.

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TMI ACTION PLAN ITEM NO. II.K.3.25 EFFECT OF LOSS OF ALTERNATING CURRENT POWER ON PUMP SEALS NRC POSITION The licensees should determine, on a plant-specific basis, by analysis or experiment, the consequences of a loss of cooling water to the reactor recirculation pump seal coolers.

The pump seals should be designed to withstand a complete loss of alternating-current (ac) power for at least 2

hours.

Adequacy of the seal design should be demonstrated.

RESPONSE

As indicated in our letter of December 17, 1980, the original design of Nine Mile Point Unit 1 includes supplying emergency power to the components which provide cooling water to the reactor recirculation pump seal coolers, thus precluding damage to the seals as a result of a loss of offsite AC power.

Therefore, no analysis or further action is required on this item for Nine Mile Point Unit l.

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TMI ACTION PLAN ITEM NO. II.K.3. 27 PROVIDE COMMON REFERENCE LEVEL FOR VESSEL LEVEL INSTRUMENTATION NRC POSITION Different reference points of the various reactor vessel water level instruments may cause operator confusion.

Therefore, all level instruments should be referenced to the same point.

Either the bottom of the vessel or the top of the active fuel are reasonable reference points.

RESPONSE

Technical Specification changes which reference the existing reactor vessel water level instruments to the same point (65 inches below the minimum normal water level at elevation 302 feet 9 inches) were submitted on August 5, 1980 for NRC review and approval.

The water level instrumentation being installed to meet the requirements of TMI Action Plan Item No. II.F.2 "Instrumentation for Detection of Inadequate Core Cooling" has its reference zero at the top of the upper grid plate at elevation 291'-3/8".

The different reference points are justified because of the different utilizations of the instrumentation.

The existing instrumentation would be used by the operators during normal operation and transients, while the inadequate core cooling water level instrumentation would be utilized by the operators during accidents in which inadequate core cooling may be present.

Additional justification is also provided in a TMI BWR Owners Group report titled NUREG 0737 II.K.3.27 Common Water Level Reference which was transmitted to the Nuc ear Regu atory Commission y a etter ate December 29, 1980 from Mr. D. B. Waters, Chairman TMI BWR Owners Group to Mr. D.

G. Eisenhut.

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TMI ACTION PLAN ITEM NO. II.K.3.28 VERIFY QUALIFICATION OF ACCUMULATORS ON AUTOMATIC DEPRESSURIZATION SYSTEM VALVES NRC POSITION Safety analysis reports claim that air or nitrogen accumulators for the

" automatic depressurization system (ADS) valves are provided with sufficient capacity to cycle the valves open five times-at design pressures.

GE has also stated that the emergency core cooling (ECC) systems are designed to withstand a hostile environment and still perform their function for 100 days following an accident.

Licensee should verify that the accumulators on the ADS valves meet these requirements, even considering normal leakage.

If this cannot be demonstrated, the licensee must show that the accumulator design is stH 1 acceptable.

RESPONSE

As indicated in our letters of June 20, 1980 and December 17, 1980, this item is not applicable to Nine Mile Point Unit 1.

The Nine Mile Point Unit 1

design includes electromatic relief valves for it's Automatic Depressurization System (ADS) and not valves operated by accumulators.

Therefore, no further action on this item is required for Nine Mile Point Unit 1.

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TMI ACTION PLAN ITEM NO. II.K.3. 29 STUDY TO DEMONSTRATE PERFORMANCE OF ISOLATION CONDENSERS WITH NONCONDENSABLES NRC POSITION If natural circulation plays an 'important role in depressurizing the system (e.g.,

in the use of isolation condensers),

then the various modes of two-phase flow natural circulation, including noncondensables, which may play a significant role in plant response following a small-break loss-of-coolant accident (LOCA) should be demonstrated.

RESPONSE

As indicated in our letters of November 7,

1980, and December 17, 1980, the emergency (isolation) condensers at Nine Mile Point Unit 1 are being modified so that the tube side of the condensers can be vented to the torus under accident conditions.

Therefore, a study to demonstrate the performance of the Nine Mile Point Unit 1 emergency condensers with noncondensables is not required.

No further action is required on this item for Nine Mile Point Unit

l.

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TMI ACTION PLAN ITEM NO. II.K.3.44 EVALUATION OF ANTICIPATED TRANSIENTS WITH SINGLE FAILURE TO VERIFY NO FUEL FAILURE NRC POSITION For anticipated transients combined with the worst single failure and assuming proper operator actions, licensees should demonstrate that the core remains covered or provide analysis to show that no significant fuel damage results from core uncovery.

Transients which result from a stuck-open relief valve should be included in this category.

RESPONSE

As indicated in our December 17, 1980. letter, this was to be addressed through a generic submittal by the BWR Owners Group.

The report titled NUREG-0737 I.K.3.<<-Ad C

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D.

G. Eisenhut.

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TMI ACTION PLAN ITEM NO. II.K.3.45 EVALUATION OF DEPRESSURIZATION WITH OTHER THAN AUTOMATIC DEPRE SURIZATION SY TEM NRC POSITION Analyses to support depressurization modes other than'ull actuation of the automatic depressurization system (ADS) (e.g., early blowdown with one or two safety relief valves (SRVs) ) should be provided.

Slower depressurization would reduce the possibility of exceeding vessel integrity limits by rapid cool down.

RESPONSE

As indicated in our December 17, 1980 letter, this item was to be addressed through a generic submittal by the BWR Owners Group.

The report titled NUREG-0737 Item II.K.3.45 - Alternate Modes of Depressurization was transmitte to t e Nuc ear Regu atory omoission y a etter ated December 29, 1980 from Mr. D. B. Waters, Chairman TMI BWR Owners Group to Mr. D.

G.

Eisenhut.

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TMI ACTION PLAN ITEM NO. II.K.3.46

RESPONSE

TO LIST OF CONCERNS FROM ACRS CONSULTANT NRC POSITION GE should provide a response to the Michelson concerns as they relate to BWRs.

Licensees should access applicability and adequacy of this response to their plants.

RESPONSE

Niagara Mohawk's submittal of August 1,

1980 addressed this item.

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TMI ACTION PLAN ITEM NO. III.A.2 IMPROVING LICE NSE E EMERGE NCY RE AREDNE L NG TERM NRC POSITION Each nuclear facility shall upgrade its emergency plans to provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.

Specific criteria to meet this requirement is delineated in NUREG-0654 (FEMA-REP-l), "Criteria for Preparation and Evaluation of Radiological Emergency

Response

Plans and Preparation in Support of Nuclear Power Plants."

RESPONSE

Niagara Mohawk has upgraded the Nine Mile Point Unit 1 Emergency Plan.

The upgraded Nine Mile Point Unit 1 Emergency Plan has been submitted under separate cover by a letter dated December 30, 1980 from Mr. D. P. Disc to Mr.

H. R. Denton.

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TMI ACTION PLAN ITEM NO. III.D.3.3 IMPROVED INPLANT IODINE IN TRUMENTATION UNDER A

IDENT CONDITI NS NRC POSITION (1)

Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident.

(2)

Each applicant for a fuel-loading license to be issued prior to January 1,

1981 shall provide the equipment,

training, and procedures necessary to accurately determine the presence of airborne radioiodine in areas within the plant where plant personnel may be present during an accident.

RESPONSE

Niagara Mohawk's submittal of December 17, 1980 indicated that a new type of charcoal cartridge and the analysis procedures were being evaluated.

As indicated in a separate letter dated December 31, 1980 from Mr. D. P. Disc to Mr. D.

G. Eisenhut, this evaluation and issuance of approved procedures will not be completed until January 31, 1981.

As indicated in our letter of December 17,

1981, the modification required to accurately measure iodine will be to provide a dedicated source of outside air.

The laboratory area ventilation air supply is being modified to transfer from the discharge of the Turbine Building ventilation supply fan to a direct source of outside air on a loss of all offsite power.

This modification will be completed during the spring 1981 refueling outage as indicated in a separate letter dated December 31, 1980 from Mr. D. P. Disc to Mr. D.

G.

Eisenhut.

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CO TMI ACTION PLAN ITEM NO. III.D.3. 4 S

Position In accordance with action item III D.3.4, Control Room Habitability, licensees shall assure that control room operators will be adequately protected against the effects of accidental release of toxic and radioactive gases and that the nuclear power plant can be safely operated or shut down under design basis accident conditions (Criterion 19, "Control Room", of Appendix A, "General Design Criteria for Nuclear Power Plants", to 10CFR Part 50).

~Res onse Niagara Mohawk has reviewed the Nine Mile Point Unit 1 control room for conformance with sections 2.2.1-2.2.2, 2.2.3 and 6.4 of the Standard Review Plan.

The results of the review are presented herein.

Potential accidents involving releases of toxic substances from off-site and on-site locations were evaluated.

Procedures outlined in Appendix B to Reg.

Guide 1.78 were used to perform the evaluation.

The results of the evaluation indicate that the release of any of the identified toxic materials within a 5 mile radius would have virtually no impact on the control room atmosphere.

Off-site accidents were conservatively considered to be ground level, directly upwind, puff releases.

On site accidents were assumed to be instantaneous, stack releases except for nitrogen which is stored outdoors.

In all cases, the distances and toxicity limits involved precluded the concentrations in the control room from approaching dangerous limits.

The control room habitability was also evaluated for onsite radioactive releases.

The following assumptions were used in the evaluation:

a) b)

c) d)

Reg.

Guide 1.3 source terms and meteorology for an elevated release (130 ).

MSIV leakage equal to Technical Specification limits and containment leakage equal to design limit.

No credit for control room filtration system A delay in releases from the main steam lines to the turbines of approximately 25 hrs.

After 100 days, the total whole body integrated dose has been calculated to be approximately 0.85R.

This value includes shine effects which may contribute to the dosage in the control room and is less than the limit, as outlined in General Design Criteria 19.

The integrated thyroid dose has also been determined not to exceed the General Design Criteria 19 limit of 30R.

The above information demonstrates that the control room at Nine Mile Point Unit 1 is adequately protected from potential accidents involving toxic or radioactive releases in the vicinity of the plant.

Therefore, no modifications are proposed to the habitability systems of the control room.

Information needed for an independent analysis is provided below.

1)

Control room mode of operation - zone isolation with incoming air filtered and a positive pressure maintained by ventilation fans during accident conditions.

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Control Room Characteri sties:

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Air volume of emergency zone - approximately 133,200 cf.

Control room emergency zone - main control room, auxiliary control room, toilet, kitchen, instrument shop, shift supervisor's office.

Control room ventilation schematic - see Figure l.

Infiltation leakage rate - assumed 10 cfm.

Positive pressure is maintained in the control room.

HEPA efficiency 99 percent DOP aerosol charcoal adsorber efficiency - 90 percent methyl iodide removal Closest distance between containment and air intake-approximately 290 feet.

Layout of control room - see Figures 2, 3.

Control room shielding - 12" solid concrete blocks on north and west wall 8" concrete blocks on south wall 8 1/2" poured slab above 5 3/4" concrete below Automatic isolation capability - none.

Isolation is by operator action.

Damper Closing Time - approximately 45 sec.

Damper Area - 14" butterfly valve Damper Leakage - 0 at 40 psf differential pressure Chlorine detectors or toxic gas detectors

- none Self contained breathing apparatus

- 2 masks with at least 8

hours oxygen supply, 2 Scott air packs at 30 minutes

each, and 2

escape packs at 5 minutes each are provided in the control room.

Bottled air supply -

2 tanks, each with at least an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> supply of oxygen (see k above).

Emergency food and water -

5 day supply for 5 men readily available.

Personnel capacity - at least 5 men for 5 days Potassium iodide drug supply - supply of potassium iodide tablets available from the Supervisor of Radiation Chemistry 3)

On-site storage of chlorine and other hazardous chemicals - see Attachment 1.

4)

Off-site manufacturing,

storage, or transportation facilities of hazardous chemicals - see Attachment l.

5)

Technical Specifications:

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b.

Chlorine detection system - none.

Control room emergency filtration system - see Attachment 2.

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IDENTIFICATION OF OFF-SITE HAZARDOUS CHEMICAL SOURCES 1)

Name and Address of Company:

2)

Type of Industry:

3)

Chemicals Stored or Used at Plant:

Alcan Sheet and Plate,

Oswego, N.Y.

Melting, Casting and Rolling of Aluminum Chemical Name Carbon Dioxide (liq)

Chl orine (1 iq)

Propane (liq)

Nitrogen (liq)

Sulfuric Acid Maximum Amt. Stored at One Time 114,000 lbs 20-1 ton cylinders 100,000 gal 13,000 gal 6,000 lb Pressure and Tem erature at Which Chemical is Stor 40 atm 9 20 C

4.8 atm C~ 20o C

6.8 atm 9 20o C

13.6 atm 9 -1800 C

1 atm 9 20o C

4)

Transportation of Chemicals:

Chemical Name Mode of Trans ort Maximum guantity/

Frequency Route*

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8 Temp.

-~Di 5li Carbon Dioxide Chlorine Propane Nitrogen Sulfuric Acid Tank Trailer Flat Bed Trailer Tank Trailer Tank Trailer

-Trucks w/Caboloy Buffalo to Alcan Buffalo to Alcan

'yracuse to Alcan Buffalo to Alcan Syracuse to Alcan 40,000 lb 12-1 ton cylinders 9000 gal 5000 gal

.~6000 gal Monthly Meekly Biweekly Biweekly Biweekly 4.8 atm 9 20 C

6.8 atm "

P,l,

• Normally via Rt. 104, Kocher Road, and County Route l.

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IDENTIFICATION OF OFF-SITE HAZARDOUS CHEMICAL SOURCES 1)

Name and Address of Company:

2)

Type of Industry:

J.

A. FitzPatri ck Nuclear Power Pl ant, Scriba, W.Y.

Nuclear Power Plant 3)

Chemicals Stored or Used at Plant:

Chemical Name Nitrogen (liq)

Hydrogen Sulfuric Acid Sodium Hydroxide Carbon Dioxide Propane Maximum Amt. Stored at One Time 10,000 gal 28,800 ft3 5,000 gal 5,000 gal 26,000 lbs 1,000 gal Pressure and Tem erature at blhich Chemical is Stor 217 psi 9 -200 F

2400 psi Ambient Ambient 340 psi 9 OoF

( 50 psi 4)

Transportation of Chemicals:

Chemical Name Nitrogen Hydrogen Sulfuric Acid Sodium Hydroxide Carbon Dioxide Propane Mode of Trans ort Truck Route*

Local Maximum guantity/

Shi ment 6900 gal 128,000 ft3 3000 lbs 3000 lbs 6900 gal 22,500 lbs Frequency of Shi ment Monthly 2/Month Monthly Monthly Monthly Press.

& Temp.

. ~Oi Sh" 250 psi 9 -200o F

2400 psi NA NA 340 psi 8 0o F

(50 psi

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• Deliveries made from Syracuse-Oswego-Roch er Area.

Major routes are Route 104 and Inter 81.

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IDENTIFICATION OF ON-SITE HAZARDOUS CHEMICAL SOURCES 1)

Name and Address of Company:

2)

Type of Industry:

3)

Chemicals Stored or Used at Plant:

Nine Mile Point Unit 1, Scriba, NY Nuclear Power Plant Chemical Name Nitrogen Carbon Dioxide Sulfuric Acid Sodium Hydroxide Maximum Amt. Stored at One Time 15,000 gal 20,000 lb 3,500 gal 3,500 gal Pressure and Tem erature at >lhich Chemical is Stor I

220 psi 9 -200 F '1 )

300 psi 8

Oo F

Ambient

Ambient, 4)

Transportation of Chemicals:

Chemical Name Mode of Trans ort Route*

Maximum guanti ty/

Frequency Shi ment

~fShi Press.

8 Temp.

Durin Shi ment Nitrogen Carbon Dioxide Sulfuric Acid Sodium Hydroxide Truck Local 6000 gal 3 ton 3000 gal 40,000 lbs Once per month II 11 II Once per quarter II II II 220 psi 9 -200 F

300 psi 9

Oo F

Ambient Ambient

• Major routes are Route 104 and Interstate 81 P;3, (1) Estimated

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~.5 COliTFOL POOVE AIR TREATY.""iiT SYSTEH

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Applies to tho operating status of the con-trol room air treatm nt system.

Objccti ve:

To assure the capability of the control roa~

air treatm nt system to minimize the amount of radioactivity or other gases entc.ing the control room in thc event of an incident.

4.4.5 CONTROL ROOY AIR TREATi'<T SYSTE 1

~All bii" Applies to the testing of the control room ai r trea tm nt system.

Objecti vc:

To assure the operability of the control room air trea tmen t sys tern.

S~if S ecification:

a.

Exc pt as specified, in Spo'cification 3.4. 5e be 1 oii, the con trol room air treatm nt sy-tern and the diesel generators required for operation of this syst m shall. be operablc at all tires

< hcn containm nt integrity is required.

The results of the in-place cold 00P and halogcna:cd hydrocarbon tes t d sign fiows on ilEPA filt rs and char-coal ad".orber ban~s shall sho'il 99".~

~ HP rc;.o:al ano

> ";9." halogcnatcd hydro".

carbon ic;;.oval Mncn tested in accordance i ith AiiSI li.510-19I5.

a.

At -1 eas once per operating cycle, or once every 18 rior.ths, whi hcver occurs first, the prcssure drop across the cor.;bined HEPA filters and charcoal adsorb r banks sh ll be derionstrated to be less than 6 inches of :vater at sv-"

m.

design -lo:; rate

(-: 10:!).

b ~

T.i'. tests aiid sample anai." is o f S~'cification 3

4 5b, c and d shgll bc perforo;cd at least once pei operating cycle or oi'.ce every 18 montlis, or after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system opel ation, 'Ivhichever occurs first or folloiving sionificant p :nl:in', fire or chemical i-cl"ase in c'ly v n i 1 a tio.i zo>>e co:-::.;uiiica t i i:g >I'li sys l.cm.

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The results of laboratory carbon sample analysis shall snow

> 90/ radioactive rethyl iodide removal when tested in accord-ance with AiiSI,'l.510-1975 at 130C and 95'.H.

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Fans shall be sho in to operate wi thin

~ 108 design flo'" when tested in accordance wi th Ai)S I l(.510-1975.

e.

From and after rhe date that the control rco,",. air trcatr.:ent system is rade or found to be ino"crable for any reason,

. reactor opera t ion or re fuel ing op"rations is per;:.issible only during the succe ding sev n days unless the system is scca>er made operabl".

c.

Cold tLP testing shall be perforired a ter each co;~piete or partial replacer'."nt of the H=PA filter bank or after any struc-tural n'aintcnanc on the system housing.

d.

)(alogenatcd hydrocarbon testing shall be perron'ed after each comolctc or partial replacement of the charcoal adsorber bank or after any structural maintenance on thc system housing.

e.

The system shall be-operated at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month.

If these cendi tions cannot bn r..ct, reactcr shutdown shall be initiated and tho reactor shall be in cold shutdown within

"-6 hours for react. or operations and re-fueling op ra"ions shall be tcr;.:inated L'l1thin 2 nours.

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ATTACHMENT 1 TO TMI ACTION PLAN ITEM NO. I.A.1.1 SHIFT TECHNICAL ADVISOR CURRENT TRAINING PROGRAM Assistant Station Shift Supervisors (Shift Technical Advisors) trainees were enrolled in the Nine Mile Point Senior Reactor Operator License Preparation Program (outline attached).

This program was modified to include training in the Shift Technical Advisor Function and to include those areas of training that were anticipated to be new requirements (i.e. thermodynamics, fluid flow, and increased training in transients and accidents).

In addition, the Assistant Station Shift Supervisors were given an individual Nine Mile Point Training Manual to direct and document their on-the-job training received while covering shift responsibilities.

The Senior Reactor Operator License Preparation Program is a

17 week program that has been used at Nine Mile Point for the past several years.

It includes 15 weeks of formal classroom training in plant specific items (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week),

one

( 1) week of Simulator training and a one (1) week review period.

At the conclusion of the training

program, a Reactor Operator written examination, Senior Reactor Operator written examination, and a walk-through oral evaluation is performed by a consultant organization, to assure candidate readiness to be presented to the NRC for licensing.

One

( 1) Assistant Station Shift Supervisor, who met the NRC eligibility requirements, was presented to the Operator Licensing Branch examiner in May 1980.

This individual passed the Senior Reactor Operator Licensing examination and received his SRO License.

It should be noted that the simulator experience provided to the five (5) Assistant Station Shift Supervisors who were not presented to the NRC for licensing was directed toward more emphasis on Accidents and Transients versus the requirements for Hot Licensing outlined in NUREG-0094, Appendix F, Paragraph D.

Other aspects of the program were identical.

At the conclusion of the training, the training documentation, including schedules, lesson

plans, quizzes,

exams, simulator experience reports, RO Written Exams, SRO Written Exams and Walk-through evaluations, wer e reviewed by the Vice President Nuclear Generation.

Based on this review, certification was issued attesting to qualification to fill the Accident Assessment Function of the Assistant Station Shift Supervisor.

The current Assistant Shift Supervisors have received the training outlined above.

A new Assistant Shift Supervisor trainee would receive this same or equivalent training.

Page 1 of 20

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II.

REQUALIFICATION All Assistant Station Shift Supervisors will attend the Requalification Simulator Program as part of the Licensed Operator Requalification Program.

This three day, 8 hour/day, program is designed to meet the requirements of Enclosure 4, Control Manipulations of Mr. H. R. Denton's.

March 28th, 1980 letter.

Assistant Station Shift Supervisors will as a

minimum perform the role of an Accident Assessment Technical Advisor.

It is anticipated that the Assistant Station Shift Supervisor will be enrolled in the NRC approved Licensed Operator Requalification Program.

Participation in this program would be identical to that of an NRC Senior Reactor Operator License holder.

III.

LONG TERM TRAINING FOR ASSISTANT STATION SHIFT SUPERVISORS It is intended that all Assistant Station Shift Supervisors obtain NRC Senior Reactor Operator Licenses as they meet the eligibility requirements.

IV.

NIAGARA MOHAWK'S TRAINING PROGRAM AS IT PERTAINS TO THE INPO DOCUMENT EN I NU N

AL AD M NDATIONS FOR P

I N DE TI N

QU LIFICATION, E U ATI N AND TRA NING" RE

. 0, A RIL INPO SECTION 5.1 Education 5 Training 5.2 Experi ence 5.3 Absences from Duties 6.1.1 Education 6.1.2 College Level Fundamental Education COMPARISON See Section 6 below All Assistant Station Shift Supervisors have a minimum of one year Nuclear Power Plant Experience, including training at Nine Mile Point Unit ¹l.

Absences from Assistant Station Shift Supervisor's duties are in accordance with the approved requalification program.

All Assistant Station Shift Supervisors have a bachelor's degree in a scientific or engineering discipline.

All Assistant Station Shift Supervisors have a bachelor's degree in a scientific or engineering discipline, plus demonstrated knowledge 1 evel equivalent to a

SRO in Reactor Theory, Reactor Chemistry, Nuclear Materials, Thermal Sciences, Electrical Sciences, Nuclear Instrumentation, and Nuclear Radiation Protection and Health Physics.

Page 2 of 20

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6.2 Applied Fundamentals 6.3 Management/Supervisory Ski 1 1 s 6.4 Plant Systems 6.5 Administrative Controls 6.6 General Operating Procedures 6.7 Transient/Accident Analysis and Emergency Procedures 6.8 Simulator Training 6.9 Annual Requalification Training All Assistant Station Shift Supervisors have a bachelor's degree in a scientific or engineering discipline and demonstrated a knowledge level equivalent to a Senior Reactor Oper'ator.

All Assistant Station Shift Supervisor s are enrolled in a Corporate Management Development Program.

Additional courses in Stress Management and Command Responsibilities and Limits have been attended or are scheduled to be attended. by the Assistant Station Shift Supervisors.

All Assistant Station Shift Supervisors have demonstrated a knowledge level equivalent to a Senior Reactor Operator.

All Assistant Station Shift Supervisors have demonstrated a knowledge level equivalent to a Senior Reactor Operator.

All Assistant Station Shift Supervisors have demonstrated a knowledge level equivalent to a Senior Reactor Operator.

All Assistant Station Shift Supervisors have demonstrated a knowledge level equivalent to a Senior Reactor Operator.

All Assistant Station Shift Supervisors have participated in Simulator Training as part of initial training.

In addition, all Assistant Station Shift Supervisors have attended or are scheduled to attend the Requalification Simulator Program designed to meet the requirements of Enclosure 4 of Mr.

H. R. Oenton's March 28, 1980 letter.

It is expected that all Assistant Station Shift Supervisors will be enrolled in the Operator Requalification Program and will participate as Senior Reactor Operator License holders.

All demonstrated knowledge is documented by quizzes and examinations.

The final examination is similar in scope and content to a NRC Senior Operator Licensing Examination.

Page 3 of 20

I r 'I NI

NIAGARA [10HAWK POWER CORPORATION NINE NI t E POINT SHIFT TECHNICAL ADVISOR TRAINING PROGRAN MEEK 0 l INSTRUCTOR(S)

MORNING Introduction to STA Function (accident assessment) o Introduction to Course o

Scope of Course o

Policies o

Lessons Learned Task Force - NUREG 0578 Introduction to BlfR's and STEAM P015ER PLN o

Reactor Core o

Steam System o

Turbine Generator o

Feedwater o

Reactivity Control o

Feedwater Control Station Reference Mater-ials in Control Room o

FSAR o

Technical Specs.

o Operating Procedures o

Special Procedures o

Admin. Procedures o

P. ID's Station Reference Mater-ials (Continued) o Electrical Drawings o

Print Reading o

Symbols o

Logic Diagrams WEDNESDAY NRC Function o

Atomic Energy Act o

AEC o

NRC Responsibilities o

Commissioners o

NRR o

OLB o

IE Other Advisory and Regulatory Bodies o

ANSI, ASME, ANS o

ACRS, INPO, NSAC Codes and Standards o

CFR o

ANSI Standards o

IE Bulletins o

NUREGS, Reg.

Guides THURSDAY Conduct of Station Operations o

Station Organization o

Shift Organization.

o 19atch Relief Proc.

o Selected Admin. Proc.

Conduct of Station Operations (Continued) o Selected Admin. Proc. '

Selected SOP's AY TMI-2 Lessons Learned o

B 0

10 Plant o

Scenario of TMI-2 Accident o

Davis-Besse Transient TMI-2 Lessons Learned (Continued) o Lessons Learned o

NUREG 0578 o

NUREG 0585 o

Kemeny Report o

Rogovin Report o

NUREG 0660 tC.

Rl ~

4 Page 4 of 20 LL

iS I AGARA f10HAWK'..POWER 'CORPORATION NINE NILE POINT SHIFT TECHNICAL ADVISOR TRAINING PROGRAM MEEK ¹ 2

INSTRUCTOR(S)

MORN'ING Examination Review and Critique of Examination o

Atomic and Nuclear Structure o

Radioactive Decay and Nuclear Reaction o

Binding Energy and the Fission Process o

Cross Sections, Flux and Reaction Rates o

Neutron Travel and Neutron Sources o

Neutron Mult. and the 6-Factor Formula o

Subcrit'ical haul t.

HEDNESDAY o

Reactivity, Shutdown margin and Excess Reactivity o

Prompt and Delayed Neutron Fraction THURSDAY o

Reactor Period o

Reactivity Coefficients o

Control Rod l(orth o

Fission Product

. Poisons and Samarium AY o

Xenon o

Time-in-LiFe Fffects o

Exam Review

,4 0

A

~ s

~f:

Pu~k Page 5 of 20 I, j L'"'l.!.f,

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NIAGARA f10HAHK,;,POWER CORPORATION NINE MILE POINT SHIFT TECHNI CAL ADVISOR TRAINING PROGRAM WEEK ¹ INSTRUCTOR (S) 'kens MORNING o

Examination o

Review and Critique of Examination o

Reactor Vessel and Internals (cont'd) o Vessel Instrumentation o

Reactor Vessel and Internals o

Nuclear Fuel and Core Components o

Control Rods o

CRD Mechanism o'RD Hydraulics WEDNESDAY o

RMCS o

R1%

o Reactor Recirculation System o

Recirculation Flow Control o

Liquid Poison THURSDAY o

Main Steam and Extraction Systems o

Principles of Detector Operation o

SRM AY o

IRM o

LPRM '

APRM o

TIP o

RPS o

Exam Review

, ~

EEK 0 4,

CT Systems and ECCS NDRNING INSTRUCTOR(S)

N1ACARA HOHAHK.,POh'ER CORPORAT ION ttrNE Hrr E PorNT SHI fT TECHNI CAL f>DVI SOR TRAI NING PROGRAM V.;

~ 4 1

\\

Examination Review of Examination CT Isolation

, t,1

• 4..

4

~

~ >

Primary CT Secondary CT Emergency Vent LOCA Scenario Reactor Building Vent

~\\

ADS WEDNESDAY Core Spray CT Spray THURSDAY Emergency Cooling Shutdown Cooling Head Spray AY RNCU Review and Operational Summary 4lVA44 A4

.. 4 4

~,

A

~

il.

Page 7.,of 20,. ~

~

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4 AW 4 8 t +

l '0 Al'44 448 1A (

24444, 1

484 1

P2 14

HIACARA llOHAHK..I0 ER

".ORPGRAT ION NINE t<It E POI: T SHIFT TECHNI CAL ADVI SOR TRAINING PROGRAN L';

EEK 0 5,

BOP Systems INSTRUCTOR(S)

"'"'ORN I NG Examination Review of Examination Review of h/S A V

~I RBCLC OPI Auxiliary Steam Condensate Feedwater MEDNESDAY HPCI Feedwater Control THURSDAY Circulating lYater Condenser Air Removal

- Steam Cycle Summary Off Gas Operational Summary Service 19ater Review I ~

lh ~

Paae, 8 of'28.',

GALS'j'"

i) I AGARA l10HAWK I'OWER CORPORATION NI NE NILE POINT SHIFT TECHNI CAL ADVISOR TRAI NING PROGRAM

, EEK g 6

Electricity Fundamentals INSTRUCTOR(S)

HORN I NG Examination Review of Exam DC Theory (Lessons 1 6 2)

DC Theory (Lesson 1)

Elect.

Measurements (Lesson 3)

AC (Lesson 4)

WEDNESDAY AC (Lesson 4)

SemiCond.

Basics (Lesson 5)

THURSDAY Diodes<6 Power Supplies (Lesson 6)

Temp.

Measurements and X ducers (Lesson ll)

IDAY Temp.

Measurements and X ducers (Lesson ll)

Logic 6 Logic Gates (Lesson 12)

Logic 6 Logic Gates (Lesson 12)

Review Page 9 ol" 20

J g'sf)L4v

il I AGARA llOHAWK I'OWER

(.ORPORAT ION NINE NI LE POINT SHIFT TECHNICAL ADVISOR TRAINING PROGRAM EEK g 7

Main Turbine and Electrical Systems INSTRUCTOR(S)

E MORNING Examination Review Generators AF Transformers Motors 3$ power Main Gen.

H Cooling H2 Seal Oil Stator Cooling WEDNESDAY Main Turb.

Turbine Aux.

Main Turb. Control Turb. Control Oil THURSDAY XBCLC "

Service Water Diesel Gen.

Plant Electrical Plant Electrical Review IDAY Paae 10 of 20

~ I II~

if I AGARA IIOHAWK I OWER LORPORAT ION NINE MIi F POINT SHIFT TECHNI CAL 'ADVISOR TRAINING PROGRAN EEK II 8 BOP Systems and Operations HORN I NG AF INSTRUCTOR(S)

Examination Review of Examination Fuel Pool Cooling Fuel Pool Clean-up Fuel Pool OP6 and 20 Plant Air Systems Startup Procedure OP43 Shutdown Procedure HEDNESDAY OP 43 Control Room Familiarization THURSDAY Integrated Operations Fuel Handling OP50'DAY Technical Specifications and BASES Selected Review Exam Review I

Paae 11'of 20

Ij

il I AGARA MOHAWK i'O'WER

(.ORPORAT ION N I NE NILE POINT SHIFT TECHNICAL ADVISOR TRAINING PROGRAN

, EEK /f 9 Heat Transfer Fundamentals NOR N I NG Examination and Review Fundamental Conce ts PE, KE, U, H

Definitions Properties of a Substance Temperature and the Ideal Gas INSTRUCTOR(S)

Work and Heat, Introduction to PROBLEMS The 1st Law of Thermo P-V diagrams Energy as a property 1st Law for Closed System Internal Energy 1st Law for an Open System Enthalpy Conservation of mass and the Continuity Equation Steady Flow PROBLEMS HEDNESDAY The 2nd Law of Thexmo Heat Engines and Heat Pumps Statement of 2nd Law Reversible Processes Carnot Cycle Entropy as a property T-S diagrams Entropy changes in revers.

ible processes Entropy and Lost Work PROBLEMS THURSDAY Va or Power C cle

~

'ankine Cycle Effect of P and T on Rankine Cycle Reheat Cycle The Regenerative Cycle Deviation of Real Cycles from Ideal Cycles PROBLEMS AY Fluid Flow 1-D Flow Continuity Equation Bernoulli's Equation PROBLEMS REVIEW Page. 12,.of'-20

,i 4l

'INE NII E POINT SHIFT TECHNI CAL ADVISOR TRAINING PROGRAM g

10 Fluids and Thermal Hydraulics INSTRUCTOR(S)

HORNING Examination Examination Review AF Fluid Flow Heat Transfer Fundamentals BWR Heat Transfer BWR Thermal Hydraulics Friction Pressure Drop Orificing Quality and Void Fraction Acceleration Pressure Drop HEDNESDAY Critical Power Transition Boiling Critical Quality GEXL Correlation LHGR Peaking Factors MAPRAT THURSDAY APLHGR Heat B'alance BWR Heat Balanqe I DAY Problems in Heat Transfer and Thermal Hydraulics Review Page 13,of:;20

I (

e

<c i ricisiisis: lutlai~~ri a><utt '~;I Ul(R I LUI'0 N I NE MI LE PO I NT SHIFT TECHNI CAL ADYI SOR TRAINING PROG RAN INSTRUCTOR (S) >~"=<<Po'-:---"

EEK 0 ll Materials and Process Instrumentation MORNING Examination Examination Review Fracture Modes Reactor Materials Neutron Embrittlement Corrosion M-W Reactions Temperature Reactor Water Chemistry Thermocouples Temperature Pressure WEDNESDAY RTD's Thermistors Manometers Elastic Deformation Elements Level Level THURSDAY Direct Methods Compensation Inferred Methods

,Techniques I

AY Level Compensation Techniques Flow Heat Type Area Type Flow Heat Type Area Type

~ Extraction eC Page 14,of 20

'i i '"~ii -". '

ll!

'IAGARA f10HAWK POWER CORPORATION NINE MILE POINT SHIFT TECHNI CAL ADVISOR TRAINING PROGRAM 4 i '&)LEi'.'. si' i<I.'EEK

¹

]2 INSTRUCTOR(S) 5/12/80 MORNING Introduction to Transient Transients:

'/U of Cold Recir Loop Recirc Pump Trips AF Transients (cont'd)

'Recirc Pump Stall

'Flow Controller Mal-functions 5/13/80 Transients (cont'd)

'nadvertant Relief Valve Actuation

'Safety Valve Actuation Transients (cont'd)

'SIV Closure Feedwater Controller'alfunction e

HEDNESDAY 5/14/80 Transients (cont'd)

~ Turbine Trips 1)

Low Power/High Power

2) Bypass/No Bypass Transients (cont'd)

Inadvertant Opening of one bypass valve Pressure Regulator Malfunction THURSDAY 5/15/80 Accidents Intro. to Accident Anal-ysis

~ Main Steam Line Break Accidents (cont'd)

'od Drop Refueling Accident LOCA's DAY 5/16/80 Accidents (cont'd)

Containment DBA TMI Summary Examination

'IC

-hI Page 15 of 20 la

S

'tf t,lk k

e

~ s%ka~w z: '""

NIAGARA flOHAWK POWER CORPORATION NINE NILE POINT SHIFT TECHNI CAL ADVISOR TRAINING PROGRAM MEEK ff 13 Health Physics 5 Rad. Prot.

MORN I NG AF INSTRUCTOR(S)

Principles of Radiation Detection Health Physics Fundamentals 5/19/80 5/20/80 Health Physics Fundamentals Health Physics Fundamentals NEDNESDAY

-5/21/80 Health Physics

.Fundamentals RPP's 10 CFR 20 10 CFR 100 THURSDAY 5/22/80 Plume Models Gaussian Plume EPA 520:

Correlation of Concen-trations with Dose Rates DAY Derivation of Dose Rate from Activity Review 5/23/80 Page 16 of 20 Pg

g~IIQ>f.

2:" 2" a~

IIIAGARA f10HAWK POWER CORPORATION NiNE MiLE POINT SHIFT TECHNI CAL ADVISOR TRAINING PROGRAM HEEK // 14 SIMULATOR TRAINING INSTRUCTOR(S) "

5/26/80 MORN I NG T.CD Orientation Control Room Tour Rx Start-up Heat-up Rate Control Moderator Temp. Effects AF Power Escalation Turbine Roll Turbine Trip less than 30'ower Turbine Trip greater than 30~o power 5/27/80 Rod North Considerations during Start-up Pressure Regulator Failure Power Increase to 50~o Power Increase to Rated Conditions Power Range Transient OPs Hot,Peak Xenon Start-up HEDNESDAY 5/28/80 Plant Transients for Reco nition 5 Dia osis

'Turbine Trip

'Load Reject Loss of Vacuum

'SIV Failure

'ressure Regulator Failure Loss of Off-site Power THURSDAY 5/29/80 Transients for Reco nition Loss of Feedwater Heating

'Inadvertant Pump Start

'ontinuous Ro'd Nith-drawal

1) at power

2) during start-up

'ecxrc Pump Trap Loss of all Recirc.

Pumps DAY 5/30/80 Failure of Feedwater Control System LOCA:

1) small

2) intermediate

3) large Main Steam Line Break TMI Scenario for BÃR's Page 17 of 20

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lw"

".:.'~".~4).:.:

>'IAGARA I'lOHAWK POWER CORPORATION NINE MILE POINT SHIFT TECHNICAL ADVISOR TRAINING PROGRAM WEEK 8 15 EMERGENCY PLAN 5 PROCEDURES INSTRUCTOR(S)

K SF MORNING AF 6/2/80 Emergency Plan Appendix A, B, C, D

Reactor Theory Review Fundamentals of Reactor Operation 6/3/80 EPP-1 Radiation Emerg.

EPP-2 Fire Fighting Problems in Rx Theory Reactor Kinetics WEDNESDAY 6/4/80 EPP-3 Search 5 Rescue EPP-4 Contaminated Injury EPP-5 Personnel Account-ability Source Neutrons and Sub-critical Multiplication THURSDAY 6/5/80 EPP 6

5 7 Surveys EPP-8 Off-site Dose Esti-mation 4

Reactivity and Neutron Multiplication Fission Product Poisons DAY 6/6/80 Emergency Procedures Review of Bl<R Heat Trans-fer Examination Paae 18.of '20

'" 'i.':!i~~~Mi.~ '"

mi'JIAGARA f10HAWK POWER CORPORATION NI NE MILE PO I NT SHIFT TECHNICAL ADVISOR TRAINING PROGRAM WEEK ¹ INSTRUCTOR(S) 6/9/80 MORNING Simulator Review Press.

Reg. Fail..

Heatup Rate Control POAH Mod. Temp. Effects AF Simulator Review Hot Pk Xe S/U MSIV Failure LOCA 6/10/80 IE Bulletins 79-12 79-13 79-26 IE Bulletins 79-15 80-01 79-16 70-13 WEDNESDAY 6/11/80 IE Info Notices 79-13 79-37 80-01 IE Info.Bulletins 80-02 80-04 80-06 80-22 THURSDAY 6/12/80 IE Circulars 79-07 79-24 80-08 Tech.

Spec.

Review DAY 6/13/80 Tech.

Spec.

Review Review

'Exam Page 19'of 20 a~aW

S

0 INSTRUCTOR(S) gEEK g'17 Certification ML)k~ ~~~

~lJ c4 &~l,g ilIAGARA fMOHAWK POWER CORPORATION NINE Nits POINT SHIFT TECHNICAL ADVISOR TRAINING PROGRAM MORNING AF SELECTED REVIEW SELECTED REVIEW 6/16/80 6/17/80 SELECTED REVIEW SELECTED REVIEW WEDNESDAY 6/18/80 SELECTED REVIEW SELECTED REVIEW THURSDAY EXAMIN I'ION 6/19/80 REACTOR OPERATOR DAY 6/20/80 EXAMI ATION SENIOR REA R

OPERATOR

., Paqe 20 of 20

Nim Point Unit 1, PRIMARY CONTAINMENT SOLATION SYSTEH DATA PAGE 6 or 19:.-

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Positions CP 5-5-W CPv-P.r 0 itJ Ct.tJ Comp 1 iance with

. GDC-.j,",

-'5, 56 ant 57 X-139 X-139 X-140 X-140 12 12 Reactor Hater Sampl N

6 Recircu Cont..

S ra Inlet Y

1 Cont.

S ra Inlet Y.

1 80-17 80-N/A 0

FF GB

~NR CJC y6 N/

HF

/A N/A N/A N/A N/A

/A N/A N/A (8)

-149 X-1 9 X-150 12 12 12 Cont.-S Con S ra Inlet Y

1 Cont.

S ra Inl t Y

.80=33 80-36 Manual 6 0

CK Gl 60 A

AO RMC NARF ANANA N A NA ANA 0

0 0

X-150 X-156 X-156 x-I37 X-168 X-174 X-174 12 34 3

Cont.

Spray Inlet Recirc.

Pum Coolln

.Water Return ec)rculation Pum Cooling Water Sup Dr Level Imp. Line Control Rod Drive to Reactor Control Rod Drive to Reactor Y

1 N

1 A

80-35 70-92 70 93 301-114 113 Manual NlA Manual I 'B 0

GB Manual

.I N/A 0

CK Manual 0

NA I

GT MO RMC N/

30.

DC SO RM l A N A DC N

RF ANANA N

M lAt A W A QNA 0

0 C

0 0

0 NA ANA N A N A N A N A AI N A N A

.X-174 X-230 X-230'.

3/4 3/4 Control Rod,Drive N

1 to Reactor N

Pur e to TEP N

Purge to TIP 30 201.2-201.2-39 NA 0

GT X-30 3 4 WZ Pur e to TIP.

Breathing A

or rynefP N

8 201.2-40

~NA Q6 N/A 0

CK GT GT GT N

RF A WANA N/

M NgA NgA H

N/

M A N~A H

NJ M

A N.A H

N A NANANA C

g C

0 C

10

'ine oint Unit 1

PRIMARY CONTAINMENT OLATION SYSTEM DATA PAGE 11 of Ig'"

Isolation Valves O

r rt5 5-5

~ 4) Ql Cl M QJ

'D CL ~

QJ r(5 C Vl rt5 Q

'E QP Ql 5

~ Q)

Vl(V Nr Vli

~-~

Ql 5

Ql I

U Vl QlM Q'r O 0 5-r G.U 5-Q5 Ql ClE lt5 R O

it5 Tt5~

5:QP O QlQ Vl 'r 0 O

rt5 O

5-O O

l eQ. Q)O

~D O

Mr CL~

Q5 5-Q Q5 A

r QP r

r U. 5-.

QJO 5-O CJ r U O

~r-5 Q

O I

r

'r 5

Vl Q O r O

Positions QJQ H 'r Vl Q5 5-5 Qlr O rt5 O U Compl i,ance w. tl; (gC-,

.55, 56 and 57 XS-316 S-316 XS-316 XS-321 XS-321 XS-321 3/4 D/P Points for ILRT 3 4 D

P Point f r RT 3 4

-D P Points for ILRT 3 4 upp.

tmos.

1 Sam le n

Sample Su

.Am Sam le Su r.

0 Sam

. Ret 2

A 2

A A

2 A

2 A

NL2=

.23 201.2-24 N A N A 0

GT GB B

RF N

N A H

DC N

NANA D

0 C-0" 0

XS-321 XS-321'S-328 Su

r. 0 Sam

. Ret 2

S~u

r. Samole Ret.

Su

r. 0 Sam

. Ret 2

A 01.2-A

01. 2-7 N A AO N

N N A N A N A O'A N A XS-328 XS-328 XS-328 XS-824 XS-325 XS-326 XS-326 3/4 Su r.

Sam le Ret.

34 S

r.

m

~3 4 Su r.

Sam le Su l

12 Cont.

S ra Suction

'12 Cont.

S ra Suction 3

Torus S ra 3 'Torus S ra~

-Y 2

A 1

W 1

.W 109 201.2-112 80-39.

E Nanua3 Muual A

0 GB AO 0

0 N

70 AC D

0 0

0 AI

'ine Point Unit 1

PRIMARY CONTAINMENT SOLATION SYSTEM DATA PAGE 12 of 19'.

Isolation Valves

~ 0 I

4 e S-S

~ ~

QJ CJ M e-QJ 3-CL M QJN Il/l QP I

QJ lQ C VP.

EQ QJ QJ S-

~ QP VP QP Nr U VP C QJ Cl U'r-0 D O U S-QJ QJOE t5 0

~l~ ~ VP lQ C CJ 0 UK3 VP~r 0 0

'I U0 S-0 U

I r2J U

W QP

~0 IC

~0 4r Cl.M S

0 QJ S-VP U O CP sVP r

r QP

~E g'r U. J 30 std r U O

~r S

C CQV 0

I

'I

~r C

VP 0 Q r D

0 Cll QJ 4J

~ r VP U 0 U O

Positions QJ S-S QPr 0 cQ O U Compliance with GDC;I-e 55,*56 and=57 XS-327 20 XS-3 7

20 XS-327 3

XS-327 3

Suppr.

Vent&Purge N

1 Su

r. Vent&Pur e H

N Makeu

& Bleed H

Makeu

& Bleed H

N2 201-16 201-1 201.2-33

,201.2-06 0

AO A

RM 60 C

C (6)

XS-330 4

Torus Makeu Torus Makeu H

1 N

1

58. 1-Ol58.-

02 Manual 0

DCV 0

RMC N

60 RF XS-332 12 XS-334 6

Core S ra Suction Y

1 Core S ra Test Ret.

Y 1

81-01 40-05 Manual 0

GT XS-335 XE:336 Core S ra Test Ret Core pray Suction Y

1 Y

1 81-22 Manual 0

GT MO RMC N

90 AC C

C 0

0 AI AI S-342 12 XS-337 12 XS-333 12 S-340 20 RS-340 20 Gore S ray Suction Y

1 Core S ra Suction Y

1 Torus Air Vent&Fill A

Torus Aii Vent&Fill H

Cont.

S ra Suction Y

Cont.

Spray Suction Y

1 on

. pray est one W

W 9l=B 81-02 201-08 201-07 80-02 80-01 80-43

. MMj Ma Manual Manual N A 0

0 0

0 GT GT GT M

M R

C MO RMC N

70 70 NAtl N

N 5C AC D

D 9.

0 LC LC LC N A XS-352

. 4 Cont.S ra Test Line to Waste Dis osal r./

9

gfi I,

~g

~ ~ 5

Nine int Unit 1 PRIMARY CONTAINMENT ISOLATION SYSTEM DATA PAGE 13 of 19.

Isolation Valves

~ C

~ 0 rtP S

S-T ~

Ql QP D C E S

QP CL K QP tt-

'C Vl pj

'E QP QP S

~ QP tflQP Nr tIP C QP S

I U

QP 0 0'0 0 S-e G U 0

I tn rtP ltP~

rC QP O OPO trl 'r 0 0

ttP00 S-0 tP I

u

<<Z QP M-EC r 0 kr O M S-

'0 0

QP'lat QP S-til O 0 QP rtil r

QP r

U. I CJ S-0 S-0 6-I it) r V O r

S 0

CO C I

I r

C trl O 0'r Q

0 QP r

rfl 0 O U O- <<P:

Positions QP S-S-W QPs 0 rtP O U Compliance with GDC 55, 56 and 57 XS-354 XS-354 XS-354 XS-365 XS-365 XS-362 XS-363 XS-364 XS-315 4XS-31 20 20 30 Torus Spray Torus S ra Torus S ra eactor Water Cleanu Relief to Torus Torus Drain Torus Drain Torus Drain Vacuum Relief S stem Y

1 W

Y 1

W N

1 W

N 1

W N

1 H

N 1'

N 1

W

-N 1

A 80-65 80-68 63.-

02.

63. 1-01 121-03 121-02 121-01 68-01 N/A N A 0

0 0

CKNA CKNA n

N

~CK RF.N NA A

gF RF PF.

~N H

~ 'AN H

NANA NA H A 13 14 14 15 XXX.3 0

30 0

Vacuum Relief System acuum e se System acuum e ie System N

1'A A

N 1

N 1

6.6=O 68-08 68-05 68-03 68-09 68-04 68-10 14raneal Ma al Manual Ma ua Ma ~a 0

0 0

0 0

0 B

c7 AO AO ck AO B

AO ck AO P'CK AO B

AO ck AO ck AO RF

/A N HCN A

RF 1A P/A Q

~NA DP,'MC RF NAiNAA RF DP

.'-<C N

RF IANAA

..C M 15

PRIMARY CONTAINMENT ISOLATION SYSTEM WATER Page 14 of..1.g-Footnotes for Isolation Valve Table The valves in the emergency condenser steam supply line remain open during an accident unless there is a break in the emergency condenser line, indicated by high steam flow in the emergency condenser line or high radiation in the emergency condenser vents.

These signals automati cal.ly close the valves.

The air operated valve in the emergency condenser remains closed during accident conditions.

They are opened by high reactor pressure or low-low water level signals.

The air operated valves will then remain opened unless a break in an emergency condenser line is indicated as discussed in (1) above.

The Core Spray System is considered to be an extension of containment, therefore core spray valves 40-01, 40-09, 40-10 and 40 do not automatically isolate.

These valves open on a low reactor pressure signal in conjunction with a high drywell pressure or low-low reactor water level. If the Core Spray System.is not needed to maintain reactor vessel water level, these valves can be isolated manually.

The drywell vent and fill line consists of a 20-.inch line which penetrates primary containment.

Once outside primary containment, it branches into a 24-inch vent line and a 4-inch nitrogen supply line with each line containing two isolation valves.

There are two lines per penetration for the reactor recirculation system instrumentation.

Each line has a manual gate valve and a flow check valve outside containment.

The Containment Spray System is considered to be an extension of containment, therefore, the containment spray valves do not automatically isolate.

If the Containment Spray System is not required to mitigate the consequences of an accident, it can be manually isolated from the control room.

En ineered Safet Function NINE MILE NIT 1 PRIMARY CONTAINMENT ION SYSTEM DATA Page 15 of 19:.

ABBREVIATIONS Isolation Valve T e

Isolation Si nal Codes N=NO Y= YES Fluid D = Direct I = Indirect N = None Others stated in Table A =

Air "B =

- Sodium Pentaborate S

=

Steam W =

Water N

= Nitrogen 2

Isolation 'Val ve Location I = Inside Containment 0 = Outside Containment Others stated in Table l

Isolation Valve Actuation Mode Position Indication in Control Room AC DC H

=. Air

= AC

=

DC

= Hand Isolation Valve Actuator A

= Angle B

= Butterfly BCK = Ball Check BL

= Ball CK

= Check DCV = Diaphragm Control Valve FCV = Flow Check Valve FF

= Flow Fuse GB

= Globe GT

= Gate RV

= Relief SCV = Stop Check SV

= Solenoid VB

= Vacuum Breaker PLG = Plug Isolation Valve Power Source Code

~or Grou Parameter(s)

Sensed for Isolation Set Point units

'High Steam Flow-Main Steam Line High Radiation-Main Steam Line

< 105 psid

< 5 times normal background Low Low Low Condenser Vacuum High Temperature Main Steam Line Tunnel

> 7 inches mercury vacuum

< 200F Low Low Reactor Water Level B

High Steam Flow-Emergency Cooling System

> 5 inches Indicator Scale

< 19 psid Low Reactor Pressure

> 850 psig A

= Automatic HF

= High Flow M

= Manual OP

=. Overpressure RF

= Reverse Flow RMC = Remote Manual Control Room RM

= Remote Manual (Local)

DP '-- Differential Pressure Isolation Valve Positions AO = Air MO = Motor SO = Solenoid High Radiation Emergency Cooling System Vent

< 25 mr/hr AI =As Is C

= Closed 0

= Open

NINE MILE POINT UNIT 1 PRIMARY CONTAINMENT ISOLATION SYSTEM DATA Page 16 of 1>>'BBREYIATIONS (Continued)

Isolation Si nal Codes Continued Code

~or Garou Parameter(s)

Sensed for Isolation Set Point units Manual Low-Low Reactor Water Level High Area Temperature Low-Low Reactor Water Level High Drywell Pressure N/A

> 5 inches TIndicator Scale)

< 190F for Cleanup System

< 170F for Shutdown Cooling System

> 5 inches TIndicator Scale)

< 3.5 psig

NINE MILE POINT UNIT 1 PRIMARY CONTAINMENT ISOLATION SYSTEM DATA Page 17 of 19'.

Fi ure Codes 1

= C-18006-C 2 = C-18014-C 3 = C-18015-C 4 = C-18017-C 5 = C-18020-C 6 = C-18041-C 7 = C-18022-C 8 = C-18578-C 9 = C-18012-C 10 = C-18002-C Sheets 1 5 2 Sheets 1

& 2 Sheet 2

Sheet 1

Sheet 2

Shhet 1

Il P

Footnotes for Compliance with GDC 55, 56 and 57 Page 18 of 19 (1) This line meets the requirements of GDC 55 regarding two isolation barri ers.

(2)

(4)

(5)

(6)

(7)

This is an essential system which is not required to isolate on a

containment isolation signal.

This does not meet the requirement of GDC 55 in that a check valve is used as the automatic isolation valve outside containment.

However, the use of check valves as the automatic isolation valve outside containment has been previously justified in Niagara Mohawk's Technical Supplement to Petition for Conversion from Provisional Operatin License to Fu

-Term Operatin License.

T ere ore, no mo i ications are required.

This line does not meet the requirements of GDC 57 in that a check valve is 'used as the automatic isolation valve outside containment.

However, the use of check valves as the automatic isolation valve outside containment has been previously justified in Niagara Nohawk's Technical Su 1ement to Petition for Conversion from Provisional Operatin License to Fu

-Term era in License.

T ere ore, no mo>>cations are required.

This line meets the requirements of GDC 57 for a closed system penetrating containment.

This line does not meet the requirements of GDC 56 in that both isolations are located outside containments.

This is justified because these lines, although initially isolated following an accident, are required to be opened for controlled venting and purging and monitoring.

Therefore, locating the valves in a less hostile environment (i.e.

outside of the drywell during a

LOCA) is considered more acceptable for s afety.

This line meets the requirements of GDC 56 regarding two isolation barri er s.

(8)

(9)

This is an instrument line which is exempt from the requirements of GDC 55, 56 and 57.

This line does not meet the requirements of GDC 55 in that there is not an inside isolation valve which is locked closed or capable of automatic isolation.

As indicated in our response to I. E.Bulletin 79-02, this line will be modified during the spring 1981 refueling outage to provide two automatic isolation valves.

(lo) This line does not meet the requirements of GDC 56 in that the inside isolation valve is not locked closed or capable of automatic isolation.

Therefore, the inside isolation valve on this line will be a locked closed valve before Nine Mile Point Unit 1 starts up from the spring 1981 refueling outage.

Page 19 of 19 (11) The torus makeup line does not meet the requirements of GDC 56 in that both isolation valves are located outside containment and one of them is a check valve.

Although the remotely operable valve is normally closed, makeup to the torus may be required during/following an accident.

Therefore, locating the valves in a less hostile environment (i.e.

outside of the drywell during a LOCA) is considered more acceptable for safety.

The check valve will prevent any reverse flow from the drywell while the remotely operable valve is opened for torus makeup.

Therefore, no modifications are required.

( 12) These valves are not considered isolation valves from a containment isolation standpoint, since the core spray and containment spray systems are considered to be extensions of containment.

These valves close on a

containment isolation signal so that water is not diverted through the test lines when the core spray and containment spray systems are required to operate.

Therefore, no modifications are required.

( 13) The reactor water cleanup relief to torus does not meet GDC 56 in that both isolation valves are located outside containment.

Since this is a

submerged line, it is not practical to have a locked closed or automatic isolation valve inside containment.

The use of check valves as the automatic isolation valve outside containment has been previously justified in Niagara Mohawk's Technical Su plement to Petition for Conversion from Provisional Operatinq License to Fu

-Term Operatin License.

ere ore, no mo >>cations are require (14) These lines do not meet GDC 56 in that there is only one manual valve located outside the containment.

However, the lines have a bolted flange at the end which provides a second isolation barrier.

Access to these lines is also located in a locked compartment.

Therefore, no modifications are required.

(15) These lines do not meet GDC 56 in that the isolation valves are all located outside containment.

However, the function of these lines are to equalize pressure between the torus and drywell and the containment and atmosphere during normal operation.

Therefore, these valves are required to be located outside containment to perform these intended functions.

No modifications are required.

CI I

f r

TAB 1 to TMI ACTION PLAN -IT

. II.E.4.2

.CONTAINHEN TION DEPENDABILITY

""Nine~Hi.1+ Point'Unit 1

PRIHAR) CONTAINMENT ISOLATION SYSTEM DATA PAGE 1 of 19 Isolation Yalves

~ O I

S S-QP QJ JD C

S-QP 5

O QJN I(/I QJ QJ tI C VI EQP QP I M QJ I/l QP'I C

Vl Q

~

0 D S

e O U S

QP QJ ClE 0

III~

~

C QJ O UIO C/I 'r 0 0

I It$

0 S-0 O

I eQ. QPo

~0 FC

'r 0 O M 0O QJ S-e VI O 0 QJ r

Vl

~

QJ g'r U. f-QPO S

S QJ 0

I rtJ r O 0

~r S-C C

~ 0 0 C I

I

~r C

Vl 0 0 'r 00 QP0 Vl O U O

Positions QJ S-S-W QPe 0 stJ CI.LJ-Compl iance with GDC =..

55, 56 atida57 X-2A X-2B X-2B

-3A X-3B X-3B X-4A 24 Hain Steam 24 Has n Steam 24 Hain Steam 24 Hain Steam 10 18 Emergency Condense Feedwater 18 Feedwater ee water eedwater 10 Emer e 10 Emer en C

d s

Y 1

S 1

S 1

S S

S 1

S I

W 1

W "se I ~

01-01 01-03 01-04

. 39=09.

39-08 31-04 31-02 31-03 31-01 N A N/A I

GB 0

A T..

0 A

0 GT 0

GB 0

CK 0

GB 0

CK HO AO AO MO MO MO A

RF RMC RF RMC RMC RMC RMC N A 10 10

~NA 60 AC DC AC N

AC 0

C N A 0

NANA C

AI C

C 0

AI AI N AHA 0

AI NA NA X-5B 10 10 10 Emerqqenc Condense Emer en'ondense Emer enc Condense Y

1 W

1 IW 39-04 39-06 B

0 DCY N/A AO X-5B X-7 X-7 X-8 X-8 X-9 X-154 X-154 10 Emer enc Condenser.

Y 14 Shutdown Cool. Rtn.

N 14 Shutdown Cool. Rtn.

N 14 Shutdown Cool.

Su N

14 Shutdown Cool.

Su N

6 React.Cleanu Su 1

N 6

React.Cleanu 5u 1

N 6

React.Cleanu

'.R n

N 6

React.cle~anu Rtn R

1 W

1 W

1 W

39-05 38-12 38-01 38-02 33-02 33-04 33-0 33-03 N A NLA 0

GT 0

K I

GT NgA 40 0

C..

0

C

)

I lt

Nine PRIMARY CONTAINMIEN PAGE 2 of l9"'-

Point Unit 1

OLATION SYSTEM DATA Isolation Valves

~ C

~ 0 r

rIJ

$- J-

~ +J e QJ M rCE QJ 0

QJN

~r QJ I

QJ II-rtJ CM E QJ QJ I-M QJ M QJ MC M

QJ 'Z7 O r 0 D r

O U QJ QJ

) Cl r

IIJ R' rl re IIJ~

C QJ 0 010 IJI 'r 0 rr C/)47 0

I it$

O0 0

O I

O WQJ M

gC rO 4r OM 0O QJ MO O QJ r

r QJ r0'r U. J QJO 0

O O

~r I-C CQ C0 C

'r-r

~r C

M Q 0

r O

0

'l3 QJ0

+ rr M O 0 O CL <<Q:

Positions QJJ-I-W QJr'I 0 rtJ O U Compliance with GDC 55 56 and 57

'-12B X-13 X-13A X-13 r X-14

~ X-14 I X-18

~I X-18

'-19 X-19 X-19 X-20 X-25 X-25 12 12 12 12 12 12 24 20 20 4

0 Dr well Cooler BCLC fr ore Spra Inlet ore S ra Inlet ore S ra Inlet ore Spra Inlet ore S ra Inlet ore Syra~ Inlet r well Vent 8

Pur e

Air S Dr well N

V Fi 1 1 Dr well-N Vent8Fill D~re 1 1 N

VentJFi 'll Cont. Atmos. Dil

~ Sample Ret.

Cont. Atmos, Di..l Sam le Ret.

Dwell Floor Dr.

Dr ell Floor Dr.

Sum Outl et N

I 2

N 2..Q 2

I A 1

W 70-95 70-94 40-01 40-09 30&

40-10 40-11 40-12 201-09 201-10 201-31 201. 2-03 2014-32 201. 7-83.-

83.1-12 N A M

Manual Manual Manual Manual anua Manual 0

0 0

0 CK G

GT DCV N A MO AO MO AO AO RF A

A RMC RMC 30 90 90 90 N A 60 60 DC D

AC D

AC 0

AC AC D

D 0

0 0

..Q 3

AI AI AI AI C

AI C

C (4)

P l

,=."." -"'

Nine Point Unit 1

PRIMARY CONTAII'NEN OLATION SYSTEH DATA PAGE 3 of

-19.>>'solation Valves

~ C0 r

0 ~

rtl S

S-T + Cl CP J3 c E Cl CL M CP I

Cl 5

IU VP CPM U r 0 D S

CL U S-Cl Cl Xl E

0 rl rtp stp~

S:CP O UPQ VP'r 0 0

I itl O0 S-0 O

l O

W CP M

EM

~r Q 4r CLM 0O Lh Cl S-Vl O O CP rVl r

Cl g or U. P-S-

0 Ilg r O 0

~r S

C CQ 0

P-r r

~r C

Vl Q O r-Q Positions Cl M.r VP 0 0 V

~ cf CP S-S-W CPrr 0 rtp CLU Comp1iarce wi U:

55, 56 =and-57

- GDC X-26 X-26

-34 5

2=34

X-34 I X-34 I ~X-35 5J

'-35

X-35 X-35 X-36(Y)'r ell E ui

. Dr.

Sum Outlet Dr we 1 1 E ui

. Dr.

Sum Outlet R

aInst.

Raact~rR cir S

s Inst.

Reactor Recir.

S s.

Inst.

Reactor Recir.

S s.

Inst.

Reactor Recir S

s Inst.

Reactor Recir.

Sys.

Inst.

Reactor Recir S

s nst.

Reactor Recir.

S s.lN nst.

React. Recir. S s Inst N

5 W

83.

09 W

83. 1-10 N A N A N/A 0

0 0

0 DCV FCV F

V AO A

HF HF RM 60 AC D

60 A

D N A N/.

NJ N/A NJA 0

C C

(7)

HF

~

I

>',Pi.": -'ine oint Uni PRIMARY CONTAINMEN OLATION SYSTEM DATA PAGE 4 of 19:.

c.Q Isolation Valves

~

C0

+ 4J rg 5

5-

~ ~

QJ QP M 5 0 A

QJ I

C/l QJ I

QJ rg 5

C/J Ig Q QJ QP 5-

~ QP CIJ QJ C/J r-crJ c crp I/J QJ 0 CJ 'r 0 'D 5 r O

LP 0

rg rg~

r 5:QP O

CPAD C/J 'r 0 0

I rCJ CJ0 5-0 CP 1

rg 0

WQP gC

~0 Qs CL~

I CJ 5 M 0

CJ QP C/J QP 5

crl CJ O QP r

C/J

~QP I

Lr 5 QJ0 5

0 QJ 0

I CJ 0

~r 5

r

'I Cn 0 0

r Q

00 QJ0

+J 'r crJ CJ O

CJ O

Positions QJ S-5-W QPr O rg CLCC Compli.ance with

.GDC 55, 56'-and 57 X-37(5.

X-37 X-37 X-43 X-43 X-43 X-28Y5

~28 X-28 X-32 X-32 X-31 5

X-31 X-31 X-30 X-30 1

React. Reci r. Sys Ins t 1

React.Recir.S sInst 1

React.Recir.S sInst 1

React.Recir.S sInst 1

React.Recir.S sInst React.Recir.SysInst React.R cir I

eact. Rec> r. Sys Inst 1

R R

1 React.Recir.S sInst React.Recir.S sInst React.Recir.S s Inst React.Recir. SysInst React. Recir. S s Inst React.Recir.S sInst React.Recir.S sInst React.Recir.S sInst React.Recir.S sInst React.Recur.S sInst 1

React.Rec>r.SysJnst 5

5 5

~

N A N A N A N A N A N A N/A 0

FCV N A ""

N 0

FCV NA HQ NAM N

N 0

FCVNA HF N

N HF 0

GT NAM N

KF HF N/

0 GT NAM N

N 0

FCV N A HF N

N 0

GT NAM N

N 0

FCV N A HF N

N 0

GTNAM N

N KF T. N/

HF N A N A N A 0

0 0

0 0

N A N

0 0

N A N

N A ANA N A ANA NANA 0

.NA N

N A 0

N A N

N A

Nine PRIMARY CONTAINMEN PAGE 5 of Jg.-

oint Unit 1

OLATION SYSTEM DATA Isolation Valves

~ C LJ 0 I

0 M eSJ S

S

~ +

QJ QP O C E QJ O

QJN I

c/J QJ I

. QJ4-C Vl Ig O

E QP QP S-M QP QCM AI-0 VlC QJ 0 0 ~

0 O S-r Q

SJ QP QJ I

eSJ 0

rl~ ~ Vl C QJ 0 OJO Vi ro 0

eQO0 S-0 0

I O

eQ: QJ ECO CLM S

0 QJ S

Vl 0 0 QP IVl IQP SJ J S-QP 0

O 0

~r S-0 C0 C

~r C

Vl 0 0 'r O

Positions QJo 4J I

Vl 0 O U CL IQ:

QJ S

S-W QPr W I O QP O

SJ Compliance wi.th GDC 55,-56 and 57"-

X-29 (S

-29

-29 X-29 X-41 5

X-41 X-41

-4 X-44 X-44 X-44 X-42 5

-42 X-42 X-42 1

React.Recir.S sInst 1

R R

1 React.Recir 1

React.Recir.S sInst 1

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