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 I P, NAME RECIPIENT AFFILIATION EISENHUTED ~ G ~ Division of Licensing

SUBJECT:

forwards documentation of TMI action items requiring 810101 submittaliper 801031 requests Sixteen oversize drawings TITLE: General CODE: ABOIS COPIES RECEIVED:LTR Distribution for after Issuance

[of ENCL 3 SIZEl Operating License encl'ISTRIBUTION NOTES!

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL ACTION: IPPOLITOE T>> 00 13 13 I

INTERNAL: D/DIREHUM FAC08 1 1 DIREDIV OF LIC 1 1 IaE 06 2 2 NRC PDR 02 1 0 11 1 0 OR ASSESS BR 10 1 0 EG FILE 01 1 1 EXTERNAL: ACRS 09 16 LPDR 03 1 1 NSIC 05 1 1 o 9QQ 08

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UMQHA~K NIAGARA MOHAWK POWER CORPORATION/300 ERIE BOULEVARD WEST, 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 Donald P. Disc Vice President Engineering Pool PEF:ja 5 Attachment

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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)'00NH Fll.f Mm Control fP S]olci6~ooo

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TABLE OF CONTENTS ITEM NO. TITLE PAGE I.A.1.1 Shift Technical Advisor I.c. 1 Short Term Accident and Procedures Review I.C.5 Feedback of Operating Experience I.C.6 Veri fy Correct Perf ormance of Operating Activities I I.',B.. 2 Plant Shielding I I.!B '.4 Training for Mitigating Core Damage II.E.4.2 Containment Isolation Dependability I I.F.1.2 Iodine/Particulate Sampling 10 II.F.1.3 Containment High Range Monitor I I.F.1.4 Containment Pressure 13 I I.F. 1. 5 Containment Water Level II.K.3.3 Reporting Safety Valve and Relief 16 Valve F a i lures and. Chall eng es I I.K.3. 13 HPCI and RCIC Initiation Levels 17 I I.K.3.14 Isolation Condenser Isolation 18 I I.K.3.15 Isolation of HPCI and RCIC 19 I I.K.3.17 ECC System Outages 20 I I.K.3.19 Interlock on Recirculation Pump 21 Loops II.K.3.21 Restart of Core Spray and LPC I 22 II.K.3.22 RCIC Suction 23 II.K.3.24 Space Cooling for HPCI and RCIC 24 II.K.3.25 Power on Pump Seals 25 I I.K.3. 27 Cordon Reference Level 26

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TABLE OF CONTENTS (Continued)

ITEM NO. TITLE PAGE II.K.3.28 Qualification of ADS Accumulators 27 II.K.3.29 Performance of Isolation Condensers 28 I I.K.3.44 Evaluation of Transients with Sing 1 e 29 Failure I I.K. 3.45 Manual Depressuri zation 30 II.K.3.46 Michelson Concerns 31 III.A.2 Emergency Preparedness 32 III.D.3.3 Inplant Radiation Monitoring 33 III.D.3.4 Control Room Habitability 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). included at the end of this report provides a description of the current training program for the Nine Mile Point Unit Shift Technical 1

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 u - erm pera in icense.

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 T R 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 I CAT 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 Mi le Point Unit to assure that at least two 1

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 design does not include HPCI and RCIC systems with pump 1

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 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. 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 includes supplying emergency power to the components which 1

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. I I .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. I I.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

I.K.3.<<-Ad was transmitte C C 1i f T i ih Si~i generic submittal by the BWR Owners Group. The report titled NUREG-0737 to t e Nuc ear Regu atory Comm)ssion y a etter dated 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.3.45 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 L I CE 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 CONDIT I 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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TM I ACTION PLAN ITEM NO. I I I.D.3. 4 CO 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) Reg. Guide 1.3 source terms and meteorology for an elevated release (130 ).

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

c) No credit for control room filtration system d) 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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2) Control Room Characteri sties:

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

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

c~ Control room ventilation schematic - see Figure l.

d. Infiltation leakage rate - assumed 10 cfm. Positive pressure is maintained in the control room.

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

g ~ Layout of control room - see Figures 2, 3.

h. 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 J ~ Chlorine detectors or toxic gas detectors - none

k. Self contained breathing apparatus - 2 masks with at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 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.

no Personnel capacity - at least 5 men for 5 days

0. 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 At tachment 1.

4) Off-site manufacturing, storage, or transportation facilities of hazardous chemicals - see Attachment l.

5) Technical Specifications:

a~ Chlorine detection system - none.

b. 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: Alcan Sheet and Plate, Oswego, N.Y.

2) Type of Industry: Melting, Casting and Rolling of Aluminum

3) Chemicals Stored or Used at Plant:

Chemical Name Maximum Amt. Stored at One Time Pressure and Tem erature at Which Chemical is Stor Carbon Dioxide (liq) 114,000 lbs 40 atm 9 20 C Chl orine (1 i q) 20-1 ton cylinders 4.8 atm C~ 20o C Propane (liq) 100,000 gal 6.8 atm 9 20o C Nitrogen (liq) 13,000 gal 13.6 atm 9 -1800 C Sulfuric Acid 6,000 lb 1 atm 9 20o C

4) Transportation of Chemicals:

Maximum guantity/ Frequency Press. 8 Temp.

Chemical Name Mode of Trans ort Route* ~Shi ment ~fsl i -~Di 5li Carbon Dioxide Tank Trailer Buffalo to Alcan 40,000 lb Monthly Chlorine Flat Bed Trailer Buffalo to Alcan 12-1 ton cylinders Meekly 4.8 atm 9 20 C Propane Tank Trailer 'yracuse to Alcan 9000 gal Biweekly 6.8 atm "

Nitrogen Tank Trailer Buffalo to Alcan 5000 gal Biweekly Sulfuric Acid -Trucks w/Caboloy Syracuse to Alcan .~6000 gal Biweekly

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

2) Type of Industry: Nuclear Power Plant

3) Chemicals Stored or Used at Plant:

Chemical Name Maximum Amt. Stored at One Time Pressure and Tem erature at blhich Chemical is Stor Nitrogen (liq) 10,000 gal 217 psi 9 -200 F Hydrogen 28,800 ft3 2400 psi Sulfuric Acid 5,000 gal Ambient Sodium Hydroxide 5,000 gal Ambient Carbon Dioxide 26,000 lbs 340 psi 9 OoF Propane 1,000 gal ( 50 psi

4) Transportation of Chemicals:

Maximum guantity/ Frequency Press. & Temp.

Chemical Name Mode of Trans ort Route* Shi ment of Shi ment . ~Oi Sh" Nitrogen Truck Local 6900 gal Monthly 250 psi 9 -200o F Hydrogen 128,000 ft3 2/Month 2400 psi Sulfuric Acid 3000 lbs Monthly NA Sodium Hydroxide 3000 lbs Monthly NA Carbon Dioxide 6900 gal 340 psi 8 0o F Propane 22,500 lbs Monthly (50 psi

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• Deliveries made from Syracuse-Oswego-Roch er Area. Major routes are Route 104 and Inter 81.

r IDENTIFICATION OF ON-SITE HAZARDOUS CHEMICAL SOURCES

1) Name and Address of Company: Nine Mile Point Unit 1, Scriba, NY

2) Type of Industry: Nuclear Power Plant

3) Chemicals Stored or Used at Plant:

Chemical Name Maximum Amt. Stored at One Time Pressure and Tem erature at >lhich Chemical is Stor I

15,000 gal 220 psi 9 -200 F '1 )

Nitrogen 20,000 lb 300 psi 8 Oo F Carbon Dioxide Sulfuric Acid 3,500 gal Ambient Sodium Hydroxide 3,500 gal Ambient,

4) Transportation of Chemicals:

Maximum guanti ty/ Frequency Press. 8 Temp.

Chemical Name Mode of Trans ort Route* Shi ment ~fShi Durin Shi ment Nitrogen Truck Local 6000 gal Once per month 220 psi 9 -200 F II II Carbon Dioxide 3 ton 11 300 psi 9 Oo F Sulfuric Acid 3000 gal Once per quarter Ambient II II II Sodium Hydroxide 40,000 lbs Ambient

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

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LIiiITIi'GCONDITIO)l FOR OPERA i ION SiJRVE I l LAIC"- REgiJ I RE'i=NT

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.5 COliTFOL POOVE AIR TREATY.""iiT SYSTEH 4.4.5 CONTROL ROOY AIR TREATi'<T SYSTE 1

~~A~ ~All bi i" Applies to tho operating status of the con- Applies to the testing of the control room trol room air treatm nt system. ai r trea tm n t system.

Objccti ve: Objecti vc:

To assure the capability of the control roa~

air treatm nt system to minimize the amount To assure the operability of the control room of radioactivity or other gases entc.ing the air trea tmen t sys tern.

control room in thc event of an incident.

S~if S ecification:

a. Exc pt as specified, in Spo'cification a. At -1 eas 3.4. 5e be oii, the con trol room air 1

once per operating cycle, or once treatm nt sy- tern and the diesel every 18 rior.ths, whi hcver occurs first, the generators required for operation prcssure drop across the cor.;bined HEPA filters of this syst m shall. be operablc at and charcoal adsorb r banks sh ll to be less than 6 inches of :vater at sv-" m .

be derionstrated all tires < hcn containm nt integrity is required. design -lo:; rate (-: 10:!).

The results of the in-place cold 00P b ~ T.i'. tests aiid sample anai ." is o f and halogcna:cd hydrocarbon tes t S~'cification 3 4 5b, c and d shgll d sign fiows on ilEPA filt rs and coal ad".orber ban~s shall sho'il >

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bc perforo;cd at least once pei operating cycle or oi'.ce every 18 montlis, or after

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HP rc;.o:al ano > ";9." halogcnatcd hydro". 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system opel ation, 'Ivhichever carbon ic;;.oval Mncn tested in accordance occurs first or folloiving sionificant i ith AiiSI li.510-19I5. p :nl:in', fire or chemical i-cl"ase in c'ly v n i a tio.i zo>>e co:-::.;uiii ca t i i:g >I'li 1

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C L1111, IHG CO'COITION FOR OP="RATIO't SURVEILLANCE RE/VI REYiKiHT II'.

The results of laboratory carbon sample c. Cold tLP testing shall be perforired a ter analysis shall snow > 90/ radioactive each co;~piete or partial replacer'."nt of rethyl iodide removal when tested in accord- the H=PA filter bank or after any struc-ance with AiiSI,'l.510-1975 at 130C and 95'.H.

tural n'aintcnanc on the system housing.

jj i ft go~ + /J~z'e>i/c d. )(alogenatcd hydrocarbon testing shall be perron'ed after each comolctc or partial

d. Fans shall be sho in to operate wi thin replacement of the charcoal adsorber bank

~ 108 design flo'" or after any structural maintenance on when tested in accordance wi th Ai)S I l(.510-1975. thc system housing.

e. From and after rhe date that the control e. The system shall be- operated at least 10 rco,",. air trcatr.:ent system hours every month.

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".

If these cendi tions cannot bn r..ct, reactcr shutdown shall be initiated and tho shall be in cold shutdown within "-6 reactor 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 Accident Assessment Function of the Assistant Station Shift Supervisor.

the 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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I I. 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 NDAT IONS FOR P I N DE TI N QU LIFICATION, E U ATI N AND TRA NING" RE . 0, A RIL INPO SECTION COMPARISON 5.1 Education 5 Training See Section 6 below 5.2 Experi ence All Assistant Station Shift Supervisors have a minimum of one year Nuclear Power Plant Experience, including training at Nine Mile Point Unit ¹l.

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

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

6.1.2 College Level Fundamental All Assistant Station Shift Supervisors Education 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.

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6.2 Applied Fundamentals 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.

6.3 Management/Supervisory All Assistant Station Shift Supervisor s Ski 1 1 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.

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

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

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

6.7 Transient/Accident Analysis All Assistant Station Shift Supervisors and Emergency Procedures have demonstrated a knowledge level equivalent to a Senior Reactor Operator.

6.8 Simulator Training 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.

6.9 Annual Requalification It is expected that all Assistant Training 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.

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

MORNING Introduction to STA Introduction to BlfR's Function (accident and STEAM P015ER PLN assessment) o Reactor Core o Introduction to o Steam System Course o Turbine Generator o Scope of Course o Feedwater o Policies o Reactivity Control o Lessons Learned Task o Feedwater Control Force - NUREG 0578 Station Reference Mater- Station Reference Mater-ials in Control Room ials (Continued) o FSAR o Electrical Drawings o Technical Specs. o Print Reading o Operating Procedures o Symbols o Special Procedures o Logic Diagrams o Admin. Procedures o P. ID's NRC Function Other Advisory and o Atomic Energy Act Regulatory Bodies o AEC o ANSI, ASME, ANS o NRC Responsibilities o ACRS, INPO, NSAC WEDNESDAY o Commissioners Codes and Standards o NRR o CFR o OLB o ANSI Standards o IE o IE Bulletins o NUREGS, Reg. Guides Conduct of Station Conduct of Station Operations Operations (Continued) o Station Organization o Selected Admin. Proc. '

o Shift Organization. Selected SOP's THURSDAY o 19atch Relief Proc.

o Selected Admin. Proc.

TMI-2 Lessons TMI-2 Lessons Learned Learned (Continued) o B 0 10 Plant o Lessons Learned o Scenario of TMI-2 o NUREG 0578 AY Accident o NUREG 0585 o Davis-Besse Transient o Kemeny Report o Rogovin Report o NUREG 0660 tC. Rl ~ 4 Page 4 ofLL 20

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

MORN'ING Examination o Atomic and Nuclear Structure Review and Critique of Examination o Radioactive Decay o Binding Energy and and Nuclear Reaction the Fission Process o Cross Sections, Flux o Neutron Travel and and Reaction Rates Neutron Sources o Neutron Mult. and o Subcrit'ical haul t.

the 6-Factor Formula HEDNESDAY o Reactivity, Shutdown o Prompt and Delayed margin and Excess Neutron Fraction Reactivity o Reactor Period o Control Rod l(orth o Reactivity Coefficients o Fission Product

. Poisons and THURSDAY Samarium o Xenon o Time-in-LiFe Fffects o Exam Review AY

~f: Pu~k I, j Page L'"'l5 of

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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 Reactor Vessel and Internals (cont'd) o Review and Critique of o Vessel Instrumentation Examination o Reactor Vessel and Internals o Nuclear Fuel and o CRD Mechanism Core Components o Control Rods o'RD Hydraulics o RMCS o Reactor Recirculation System WEDNESDAY o R1% o Recirculation Flow Control o Liquid Poison o Main Steam and o Principles of Detector Extraction Systems Operation o SRM THURSDAY o IRM o RPS o LPRM o Exam Review AY APRM o TIP

~

N1ACARA HOHAHK.,POh'ER CORPORAT ION ttrNE Hrr E PorNT SHI f T TECHNI CAL f>DV I SOR TRA I NING PROGRAM ~

V.;

41

\

EEK 0 4, CT Systems and ECCS INSTRUCTOR(S)

NDRNING Examination CT Isolation

, t,1 Review of Examination

• 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 Review and Operational AY Summary RNCU il.

4lVA44 A4 .. 4 4 ~, A ~ ~ 4 AW 4 8 t + l '0 Al'44 448 1A (

Page 24444, 1 7.,of 484 20,.

1 P2

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14

HIACARA llOHAHK..I 0 ER ".ORPGRAT ION NINE t<It E POI: T L';

SHIFT TECHNI CAL ADV I SOR TRAINING PROGRAN EEK 0 5, BOP Systems INSTRUCTOR(S)

"'"'ORN I NG Examination Review of Examination Review o f h/S AV

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

- 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 TRA I NING PROGRAM

, EEK g 6 Electricity Fundamentals INSTRUCTOR(S)

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

Review of Exam DC Theory (Lesson 1)

Elect. Measurements AC (Lesson 3) (Lesson 4)

AC SemiCond. Basics (Lesson 4) (Lesson 5)

WEDNESDAY Diodes<6 Power Supplies Temp. Measurements (Lesson 6) and X ducers (Lesson ll)

THURSDAY Temp. Measurements and Logic 6 Logic Gates X ducers (Lesson 12)

(Lesson ll)

Review IDAY Logic 6 Logic Gates (Lesson 12)

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 E

EEK g 7 Main Turbine and Electrical Systems INSTRUCTOR(S)

MORNING AF Examination Transformers Review Motors 3$ power Generators Main Gen. H2 Seal Oil H Cooling Stator Cooling Main Turb. Main Turb. Control Turbine Aux. Turb. Control Oil WEDNESDAY XBCLC " Diesel Gen.

Service Water THURSDAY Plant Electrical Plant Electrical Review IDAY Paae 10 of 20

II ~I

~

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

EEK 8 BOP HORN I NG AF Examination Fuel Pool Cooling Review of Examination Fuel Pool Clean-up Fuel Pool OP6 and 20 Plant Air Systems Startup Procedure OP43 Shutdown Procedure HEDNESDAY Control Room OP 43 Familiarization Integrated Operations Fuel Handling THURSDAY Technical Specifications Selected Review OP50'DAY and BASES 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 INSTRUCTOR(S)

NOR N I NG Examination and Review Work and Heat, Fundamental Conce ts Introduction to PE, KE, U, H Definitions Properties of a Substance PROBLEMS Temperature and the Ideal Gas The 1st Law of Thermo Internal Energy P-V diagrams 1st Law for an Open System Energy as a property Enthalpy 1st Law for Closed System Conservation of mass and the Continuity Equation Steady Flow PROBLEMS The 2nd Law of Thexmo Entropy as a property Heat Engines and Heat T-S diagrams Pumps HEDNESDAY Statement of 2nd Law Entropy changes in revers.

ible processes Reversible Processes Entropy and Lost Work Carnot Cycle PROBLEMS Va or Power

~

C cle The Regenerative Cycle

'ankine Cycle Deviation of Real Cycles from Ideal Cycles Effect of P and T on THURSDAY Rankine Cycle Reheat Cycle PROBLEMS Fluid Flow PROBLEMS 1-D Flow AY Continuity Equation Bernoulli's Equation 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 AF Examination Fluid Flow Examination Review Heat Transfer Fundamentals BWR Heat Transfer Orificing BWR Thermal Hydraulics Friction Pressure Drop Quality and Void Fraction Acceleration Pressure Drop Critical Power GEXL Correlation Transition Boiling HEDNESDAY Critical Quality LHGR Peaking Factors MAPRAT THURSDAY APLHGR Heat B'alance BWR Heat Balanqe Problems in Heat Transfer and Thermal Hydraulics Review I DAY 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 ADY I SOR TRAI NING PROG RAN EEK 0 ll Materials and Process Instrumentation INSTRUCTOR (S) >~"=<<Po'-:---"

MORNING Examination Fracture Modes Examination Review Reactor Materials Neutron Embrittlement Corrosion M-W Reactions Temperature Reactor Water Chemistry Thermocouples Temperature Pressure RTD's Manometers WEDNESDAY Elastic Deformation Thermistors Elements Level Level THURSDAY Direct Methods Compensation Inferred Methods ,Techniques Level Flow Compensation Heat Type I AY Area Type Techniques Flow Heat Type

~ Extraction Area Type Page 14,of 20 eC

'i i '"~ii -". ' ll! 4 i '&)LE i'.'. si' f10HAWK POWER CORPORATION 'IAGARA NINE MILE POINT i<I.'EEK SHIFT TECHNI CAL ADVISOR TRAINING PROGRAM

¹ ]2 INSTRUCTOR(S)

MORNING AF Introduction to Transient Transients (cont'd)

'Recirc Pump Stall Transients: 'Flow Controller Mal-

'/U of Cold Recir functions Loop 5/12/80 Recirc Pump Trips Transients (cont'd) Transients (cont'd)

'nadvertant Relief Valve 'SIV Closure Actuation Feedwater

'Safety Valve Actuation Controller'alfunction 5/13/80 e Transients (cont'd)

~ Turbine Trips Transients (cont'd)

Inadvertant Opening

1) Low Power/High Power of one bypass valve HEDNESDAY 2) Bypass/No Bypass Pressure Regulator 5/14/80 Malfunction Accidents Accidents (cont'd)

Intro. to Accident Anal-ysis

'od Drop THURSDAY Refueling Accident

~ Main Steam Line Break 5/15/80 LOCA's Accidents (cont'd) Examination Containment DBA DAY TMI Summary 5/16/80 Page 15 of 20

'IC -hI la

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

MORN I NG AF Principles of Radiation Health Physics Detection Fundamentals 5/19/80 Health Physics Health Physics Fundamentals Fundamentals 5/20/80 Health Physics RPP's NEDNESDAY .Fundamentals 10 CFR 20 10 CFR 100

-5/21/80 Plume Models EPA 520:

Gaussian Plume Correlation of Concen-THURSDAY trations with Dose Rates 5/22/80 Derivation of Dose Rate Review from Activity DAY 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) "

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

'Turbine Trip Loss of Off-site Power HEDNESDAY 'Load Reject 5/28/80 Loss of Vacuum Transients for Reco nition 'ontinuous Ro'd Nith-drawal Loss of Feedwater 1) at power THURSDAY Heating 2) during start-up 5/29/80 'Inadvertant Pump Start 'ecxrc Pump Trap Loss of all Recirc.

Pumps Failure of Feedwater Main Steam Line Break Control System TMI Scenario for BÃR's LOCA:

DAY

1) small 5/30/80 2) intermediate

3) large Page 17 of 20

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>'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 Emergency Plan Reactor Theory Review Appendix A, B, C, D Fundamentals of Reactor Operation 6/2/80 EPP-1 Radiation Emerg. Problems in Rx Theory EPP-2 Fire Fighting Reactor Kinetics 6/3/80 EPP-3 Search 5 Rescue Source Neutrons and Sub-EPP-4 Contaminated Injury critical Multiplication WEDNESDAY 6/4/80 EPP-5 Personnel Account-ability 4

EPP 6 5 7 Surveys Reactivity and Neutron EPP-8 Off-site Dose Esti- Multiplication THURSDAY Fission Product Poisons mation 6/5/80 Emergency Procedures Examination DAY Review of Bl<R Heat Trans-fer 6/6/80 Paae 18.of '20

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

f10HAWK POWER CORPORATION mi'JIAGARA NI NE MILE PO I NT SHIFT TECHNICAL ADVISOR TRAINING PROGRAM WEEK ¹ INSTRUCTOR(S)

MORNING AF Simulator Review Simulator Review Press. Reg. Fail.. Hot Pk Xe S/U Heatup Rate Control MSIV Failure POAH LOCA Mod. Temp. Effects 6/9/80 IE Bulletins IE Bulletins 79-12 79-15 79-13 80-01 79-26 79-16 6/10/80 70-13 IE Info Notices IE Info. Bulletins 79-13 80-02 79-37 80-04 WEDNESDAY 80-01 80-06 80-22 6/11/80 IE Circulars Tech. Spec. Review 79-07 79-24 80-08 THURSDAY 6/12/80 Tech. Spec. Review Review

'Exam DAY 6/13/80 Page 19'of 20 a~aW

S ilIAGARA f MOHAWK POWER CORPORATION ML)k~ ~~~ *

~lJ c4 &~l,g NINE Nits POINT 0gEEK g'17 SHIFT TECHNICAL ADVISOR TRAINING Certification PROGRAM INSTRUCTOR(S)

MORNING AF SELECTED REVIEW SELECTED REVIEW 6/16/80 6/17/80 SELECTED REVIEW SELECTED REVIEW WEDNESDAY 6/18/80 SELECTED REVIEW SELECTED REVIEW EXAMIN I'ION THURSDAY REACTOR OPERATOR 6/19/80 EXAMI ATION DAY SENIOR REA R OPERATOR 6/20/80

., Paqe 20 of 20

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

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j'uPP 30 X-50 Dr well 0 Sample N 2 A 201.2- 0 0' SU 1 27 X-50 1 2 Dr well 0 Sam le N 2 J A 201.2- 0 GT SO A RM 60 DC D 0 0 C C Suppl 28 X-51 EC Elbow Flou t r Y H .0 0 N+

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Nine Point Unit 1 PR IMARY CONTAINMEN SOLATION SYSTEM DATA PAGE 7 of 19.

Isolation Valves I

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'X-71D Reactor Vessel Inst N N/A H N 0. QA

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. -'- Nine oint Unit 1 PRIMARY CONTAINMENT OLATION SYSTEM DATA PAGE 8 of 19 Isolation Valves JR I

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Nine Point Unit 1 PRIMARY CONTAINHEN OLATION SYSTEM DATA PAGE 9 of 19 Isolation Valves I

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• PRIMARY CONTAINMENT Nine 'int Unit I OLATION SYSTEM DATA PAGE 10 of 19 Isolation Valves I I CP I(5 r I V O r55 r(J ~ r 5

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'ine oint Unit 1 PRIMARY CONTAINMENT OLATION SYSTEM DATA PAGE 11 of Ig'"

Isolation Valves QJ CJ r l U O r(5 ~ r- 5 C Vl rt5 Q5 5-QJ O Positions O Q 5- Q r 'E QP eQ. Q)

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'ine Point Unit 1 PRIMARY CONTAINMENT SOLAT ION SYSTEM DATA PAGE 12 of 19'.

Isolation Valves QJ std r I U O lQ r2J ~ r S C VP. QJ S- C C Positions

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XS-335 Core S ra Test Ret Y 1 C C AI XE:336 Core pray Suction Y 1 81-22 Manual 0 GT MO RMC N 90 AC 0 0 AI XS-337 12 Gore S ray Suction Y -

1 W 9l=B . MMj XS-333 12 Core S ra Suction Y 1 W 81-02 Ma S-340 20 Torus Air Vent&Fill A 201-08 RS-340 20 Torus Aii Vent&Fill H 201-07 0 M Cont. S ra Suction Y 80-02 Manual 0 GT M R C 70 5C D 9.

S-342 12 Cont. Spray Suction Y 1 80-01 Manual 0 GT MO RMC N 70 AC D 0 on . pray est one 80-43 N A 0 GT NAtl N N LC LC LC N A XS-352 . 4 Cont.S ra Test r./ 9 Line to Waste .

Dis osal

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

Isolation Valves I

QP it) r tt- I V O r S

'C Vl

~ C pj QP S- Positions

~ 0 u CJ S- 0

'E QP <<Z QP til O

~tfl QP S 0I 0 S-0 S- 0 QPtil r

0 C O C rtP

~D S S-QP QP QP tn M- '0 I I 0 QP QP S-Ql 0' 0 ltP~

T S- S- W QP Nr S QP r

rtP C QP ttP EC 0 r QP r C r QPs Compliance with S

C E QP CL K tIP C I U

0S-e 0 G U O OPO trl 'r 0 00 tP kr0 r

O M QP'lat r

U. I 6-0 trl O 0'r Q

rfl 0 O U O- <<P:

0 rtP O U GDC 55, 56 and 57 XS-354 Torus Spray Y 1 W 80-65 N/A 0 CKNA RF.N NA A NANA NA H A XS-354 Torus S ra XS-354 Torus S ra Y 1 W 80-68 gF XS-365 20 eactor Water Cleanu N 1 W 63.- RF 13 02.

XS-365 20 Relief to Torus N 1 W 63. 1- N A 0 CKNA PF. ~N 01 XS-362 Torus Drain N 1 H 121-03 n N H ~ 'AN H 14 XS-363 Torus Drain N 1' 121-02 0 XS-364 Torus Drain N 1 W 121-01 14 XS-315 30 Vacuum Relief -N 1 A 68-01 ~CK 15 4XS-31 S stem XXX.3 0 Vacuum Relief N 1'A 6.6=O 14raneal 0 P'CK AO RF /A N System N 1 A 68-08 Ma al 0 B AO HCN A 68-05 0 ck AO RF IANAA 30 acuum System e se 68-03 68-09 Manual ck B

AO RF 1A P/A DP,'MC ~NA Q ..C M AO Ma ua 0 c7 AO RF NAiNAA acuum e ie 68-04 0 ck AO RF 15 0 System N 1 68-10 Ma ~a B AO DP .'-<C N 0 ck AO

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.

NINE MILE NIT 1 PRIMARY CONTAINMENT ION SYSTEM DATA Page 15 of 19:.

ABBREVIATIONS En ineered Safet Function Isolation Valve T e Isolation Si nal Codes N=NO A = Angle Code Parameter(s) Sensed Set Y= YES B = Butterfly ~or Grou for Isolation Point units BCK = Ball Check Position Indication in Control Room BL = Ball 'High Steam Flow- < 105 psid CK = Check Main Steam Line D = Direct DCV = Diaphragm I = Indirect Control Valve High Radiation- < 5 times N = None FCV = Flow Check Valve Main Steam Line normal background Others stated in Table FF = Flow Fuse GB = Globe Low Reactor Pressure > 850 psig Fluid GT = Gate RV = Relief Low Low Low > 7 inches A = Air SCV = Stop Check Condenser Vacuum mercury vacuum "B = -

Sodium Pentaborate SV = Solenoid S = Steam High Temperature < 200F VB = Vacuum Breaker W = Water PLG = Plug Main Steam Line

=

N 2 Nitrogen Isolation Valve Power Source Tunnel .

Isolation 'Val ve Location

=. Air Low Low Reactor > 5 inches AC = AC Water Level Indicator I = Inside Containment DC = DC Scale 0 = Outside Containment H = Hand Others stated in Table B High Steam Flow- < 19 psid l

Emergency Cooling Isolation Valve Actuation Mode Isolation Valve Actuator System

= Automatic mr/hr A

AO = Air High Radiation < 25

= High Flow HF MO = Motor Emergency Cooling .

= Manual M

SO = Solenoid System Vent OP =. Overpressure RF = Reverse Flow RMC = Remote Manual Control Room RM = Remote Manual (Local)

DP '-- Differential Pressure Isolation Valve Positions 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 Parameter(s) Sensed Set

~or Garou for Isolation Point units Manual N/A Low-Low Reactor Water Level > 5 inches TIndicator Scale)

High Area Temperature < 190F for Cleanup System

< 170F for Shutdown Cooling System Low-Low Reactor Water Level > 5 inches TIndicator Scale)

High Drywell Pressure < 3.5 psig

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

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

Il P

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

(2) 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.

(4) 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.

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

(6) 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 af ety.

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

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

(9) 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 starts up from the spring 1981 1

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 TI ON DEPENDABILITY

""Nine~Hi.1+ Point'Unit 1 PRIHAR) CONTAINMENT ISOLATION SYSTEM DATA PAGE 1 of 19 Isolation Yalves I

QJ rtJ r tI I

~

O r

0 S-C VI QJ QP Positions

~ O I QJ EQP eQ. QP S-VI O e

O S ~C 0C N QP M QJ I 0 0 S-0 o 0 QJVl 0I I Vl ~0 r C QJ S

QJ S-QP JD

(/I QJ I/l QP'I Q ~

QP S

QJ Cl ~

III C

~QJ I

It$ FC 0 ~ QJ S

QJ ~ r C I 00 QP 0 QPe S-S- W i

Compl ance wi th C 0 D E O 'rUIO 0 0

'r 0 O g'r 0 Vl 0 'r0 Vl S-5 O QP C S O U e C/I O O M U. f-O U O

0 stJ CI.LJ- GDC =..

55, 56 atida57 X-2A 24 Hain Steam 01-01 I GB HO RMC 10 AC 0 C C AI 24 Has n Steam 1 S 01-03 0 A AO X-2B 24 Hain Steam 1 S T..

X-2B 24 Hain Steam 1 S 01-04 0 A AO A RMC 10 C C S . 39=09. MO RMC 0 AI

-3A 10 Emer e S X-3B 10 Emer en C d s Y X-3B 10 Emergency Condense 1 S 39-08 0 GT MO RMC DC X-4A 18 Feedwater 31-04 0 GB AC AI 18 Feedwater 31-02 N A 0 CK RF ~NA N N A N AHA ee water I W 31-03 0 GB MO RMC N A 60 AC 0 0 AI eedwater 1 "se W 31-01 N/A 0 CK RF NANA NA NA I ~

10 Emerqqenc Condense 39-04 N/A 10 Emer en'ondense Y 1 W 39-06 B 0 DCY AO X-5B 10 Emer enc Condense 1 IW X-5B 10 Emer enc Condenser. Y 39-05 X-7 14 Shutdown Cool. Rtn. N 40 0 C.. 0 X-7 14 Shutdown Cool. Rtn. N 38-12 N A 0 K X-8 14 Shutdown Cool. Su . N 38-01 I GT X-8 14 Shutdown Cool. Su . N 1 W 38-02 6 React.Cleanu Su 1 N 1 W 33-02 X-9 6 React.Cleanu 5u 1 N 33-04 GT X-154 6 React.Cleanu '.R n -

N 1 W 33-0 X-154 6 React.cle~anu Rtn R 33-03 NLA 0 NgA

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Nine Point Unit 1 PRIMARY CONTAINMIEN OLATION SYSTEM DATA PAGE 2 of l9"'-

Isolation Valves QJ II- I O O CM rtJ ~ r I-

~ C QJ QJ O C C Positions r

~ 0 O Q QJ E QJ WQJ MO

- 0 rIJ

$ - J-

~ r N QJ I-M QJ M QJ M rl 0 I

0 M

$ O QJ C 0 C

'r- r 0 QJ QJ J-e IIJ~

~

rCE

+J QJ M QJ QJ O r 0D

'Z7

) Cl QJ QJ re 0

C QJ 010 it$ gC 0 rO r QJ r

~ r M Q C

'l3 0 rr

+ QJr I- W Compliance with 4r O O M O 'I QJ 0

I MC $ r O U r

IIJ R' IJI 'r 0 rr C/)47 0 O OM r

0'r U. J O 0 r 0 O 0 rtJ CL <<Q: O U GDC 55 56 and 57 0

'-12B Dr well Cooler 70-95 N A CK N A RF (4)

BCLC fr 70-94 M 30 DC D X-13 12 ore Spra Inlet 40-01 Manual 90 AC D 3 AI X-13A 12 ore S ra Inlet 40-09 Manual AC 0 AI X-13 12 ore S ra Inlet 30& Manual 0 AI 12 ore Spra Inlet 40-10 Manual G 90 r X-14 12 ore S ra Inlet 40-11 anua GT MO A RMC 90 AC AI

~ X-14 12 ore Syra~ Inlet 40-12 Manual N A I X-18 24 r well Vent 8 Pur e 201-09 0

~I X-18 Air S 201-10 0 AO RMC

'-19 20 Dr well N V Fi 1 1 201-31 0 MO 60 AC D C AI X-19 20 Dr well- N Vent8Fill N I 2 N AO 201. 2-03 X-19 4 D~re 1 1 N Vent JFi 'l l 2..Q 2014-32 Cont. Atmos. Dil 2 I A 201. 7- 0 DCV AO A 0 0 C C

~ Sample Ret.

X-20 Cont. Atmos, Di..l 60 D ..Q Sam le Ret.

X-25 Dwell Floor Dr. 83.-

X-25 Dr ell Floor Dr. 1 W 83.1-Sum Outl et 12

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,=."." -"' Nine Point Unit 1 PRIMARY CONTAII'NEN OLATION SYSTEH DATA PAGE 3 of

-19.>>'solation Valves I

lg r O 0 l

Cl

~ r S Positions

~ C S- C C 0 O Q

~rtl r W CP Vl O 0 0 S- O CP VP rl 0 0 M r Vl 0r P- CP T + J3S-Cl S Cl 5 CPM S-Cl Cl rtp stp~

I S-r Cl M.r CPr S-S- W CP cCl CP U r Xl S:CP itl EM 0 r Cl ~ r C E 0 D E O UPQ 0 O ~

4rQ r O 0

Vl Q VP 0 r Comp1iarce wi U:

CL M I

U I S CL U VP'r 0 O CLM Lh g

U. P-or Q

O r-

~0 cfV CLU 0 rtp - GDC 55, 56 =and-57 X-26 ell E ui . Dr. W 83. 60 AC D (7)

Sum Outlet 09 X-26 Dr we 1 1 E ui . Dr. W 83. 1- 0 DCV AO A RM 60 A D 0 C C Sum Outlet 10

-34 5 R a Inst.

2=34 Raact~rR cir S s HF Inst.

X-34

Reactor Recir. S s. N A 0 Inst.

I X-34 Reactor Recir. S s. 0 FCV HF N A Inst.

I ~X-35 5J Reactor Recir S s N A Inst.

'-35 Reactor Recir. Sys. N/A 0 F V Inst.

X-35 Reactor Recir S s nst.

X-36(Y)'r X-35 Reactor Recir. S s.lN N/. NJ N/A NJA nst.

React. Recir. S s Inst N 5 HF

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

Isolation Valves I I QJ rg 1

rg CJ

~r 0

5

~ C 5

Ig C/J CJ QP 5

QJ 0 Positions 0 A 0 5 QJ Q QJ WQP crl CJ

+ 4J

~ 5-QP 0 0 5- 5 M O QP 0 rg 5-I C/l QP crp 0 r C/J r 'I 00 QJ

~ ~M 5

QJ CIJ QJ I/J QJ 0 rg rg~ I QP QJ 0

+J 'r 5- W S-QP QJ C/J r- CJ 'r r 5:QP rCJ

~0 0CJ gC ~ QJ QPr Compli.ance with 05 r'D O 0 0r Cn crJ CJ I crJ c C/J 'rCPAD 0

CJ 0 CP Qs QP I 0 O CJ O rg .GDC 5 0 O LP CL~ C/J Lr 5 Q O CLCC 55, 56'-and 57 X-37(5. 1 React. Reci r. Sys Ins t 0 GTNAM N N 0 0 X-37 1 React.Recir.S sInst KF T. N/

X-37 1 React.Recir.S sInst HF N A X-43 1 React.Recir.S sInst N A X-43 1 React.Recir.S sInst X-43 React.Recir.SysInst N A 0 FCV N A "" N X-28Y5 React.R cir I HQ eact. Rec> r. Sys Inst 0 FCV NA ANA

~28 1 R R NAM N N 0 N A X-28 1 React.Recir.S sInst N A 0 FCVNA HF N N N A ANA React.Recir.S sInst N A X-32 React.Recir.S s Inst N A X-32 React.Recir. SysInst 5 React. Recir. S s Inst HF X-31 5 React.Recir.S sInst 5 0 GT NAM N X-31 React.Recir.S sInst KF N A NANA N/

X-31 React.Recir.S sInst HF 0 GT NAM N N 0 0 0 .NA X-30 React.Recir.S sInst N A 0 FCV N A HF N N N A N N N A X-30 React.Recur.S sInst 5

~

0 GT NAM N N 0 0 0 N A 1 React.Rec>r.SysJnst N/A 0 FCV N A HF N N N A N A N N N A

Nine o i nt Uni t 1 PRIMARY CONTAINMEN OLATION SYSTEM DATA PAGE 5 of Jg.-

Isolation Valves

. QJ 4- I

~

O r

0 S-

~ C C Vl Ig QJ S

Positions LJ 0I O O 0 0M QJ E QP eQ: QJ Vl 0 N QP S-M QP 0 0 S- S 0 VlQP C 0

eSJ

+O S S I

c/J rl

~ ~ Vl 0 I

~ rC QJ o

QJ S

0 0~

~ QJ QCM S- S- W Compliance wi.th QP QJ AI-0 QJ QP QJ C QJ eQ EC 0 I QP QP C 4J I QPr C E Vl C 0S-r O I 0 ro OJO O 0 0 O 0 0 'r0 Vl Vl 0 WI GDC QJ I eSJ Vi O U O QP O Q SJ CLM SJ J O CL IQ: O SJ 55,-56 and 57"-

X-29 (S 1 React.Recir.S sInst 5 W 0 GTNA 0'

-29 1 R R HF

-29 1 React.Recir 'N N A X-29 1 React.Recir.S sInst 5 W HF X-41 5 1 React. R ir . GT NA X-41 1 React.Recir.S slnst W 0 FCV NA N A X-41 1 React.Recir.S sInst

-4 X-44 R~, Rmi r.SZs~

React.Recir.S sInst 0 GT QA X-44 1 React.Recir.S sInst HF X-44 1 React.Recir.S sInst N 0 N A 1 React.Recir.S slnst 0 FCV N A HF A NA NA X-42 5 React. Reci r.~Ss Inst 5 W 8

-42 R~a~ war~ 0 FCV N A HF X-42 React.Recir.S sInst Jjl&

X-42 React.Recur.S sInst 5 W N A HF X-38 React.Recir.S sInst 0 GT A N A X-38 React.Recir.S sInst O' N A 0 FCV N A HF N A N A

-38 /A.

X-38 React.Recir.S sInst 5 W HF X-47 React.Recir.S sInst N A

-47 React.Recir.S sInst 5 W N A X-47 R tR S I st 0 GT NA N A N 0 X-47 1 React. Recir. Sl s Inst N 5 W Hf

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