ML19210E580

TMI-1 Restart Rept, Amend 7

ML19210E580
Person / Time
Site:	Three Mile Island
Issue date:	12/03/1979
From:	METROPOLITAN EDISON CO.
To:
Shared Package
ML19210E576	List: ML19210E575 ML19210E580
References
NUDOCS 7912050325
Download: ML19210E580 (100)

Text

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7.3.5.2 Sannle Drains The TMI-l Nuclear Sanpi' sanple and analysis drains are ruited to the TMI-l Au::iliary # aing Sump to 'e o processed by the Liquid Radwaste Systen. In _.e event of a plant accidev. accident uater lacluding reactor coolant would be discharged to the sunp as the result of sanpling and analysis. Since the sunn is not sealed radioactivity _fron the accident would escape uncontrollably into the Au::iliary building f ron the sunp.

To prevent this f rom happening, nodifications will be nade to pipe the radiochemical laboratory drains to the Miscellaneous '.?aste Storage Tank either directly or by way of an intermediate collection tank and punp(s). The Miscellaneous Waste Strrage Tank would contain the laboratory wastet because the tanks gas space is cannected to the vent header. Intern siate tanks and nunps that nay be used would also be vented to the :aste gas syste The laboratory waste collection nodification will be operational by October 1, 1980.

7.3.3.3 Inoroved In-Plant Radiciodine Monitorin? Instrumentation Both the TMI-1 Control Room anl the Health Physics Laboratory will be provided with analytical equipment capable of distinguishing radioactive Iodine-131 fron other airborne radionuclides. Th e in-strumentation will consist of NaI crystals with a single or dual channel analyzer capable of setting low-level discrimination and window width. lodine-131 settings will be pre-establish as with other nuclides such as Rb-88 and Cs-137 to allow for short-tern identification of radiological hazards and detcrnination of the need for respiratory protection.

The equipment will be checked quarterly and r2 calibrated as necessary. Functional checks will be performed nonthly to insure instrunent operability.

1. " M l0 D'LMm *D'3'{UAL JL i494 014 7-15 A 7

QUESTION

15. Provide your evaluation of anti.:ipatory reactor trip parameters (feedwater pump turbine control oil rather than feed flow or other parameters). Include your evaluation of the need for a low steam generator level trip addressing various power levels. Discuss those transient scenarios that may not initiate anticipatory reactor trip for certain loss of feedwater/ condensate events (rather than high pressure reactor trip).

RESPONSE

Feedwater (FN) pump turbine control oil pressure was selected rather than feedwater flow since any set point based on feedwater flow that would be i anticipatory of reactor trip on high pressure would interfere with normal I operatien. FN pump control oil pressure loss on the other hand is anticipatory l of loss of FN and does not interfere with normal operation. The only LOFW [

that could occur and not resul t in an anticipatory reactor trip (ART) is  !

inadvertent simultaneous closure of both FN control valves (the FN pumps trip on loss of condensate / condensate boostcr pump pressure at the FW pump suction thus an ARTS would occur).

The primary purpose of anticipatory reactor trips (ARTS) is to reduce the probability of lifting the 3RV for turbine trip / loss of main feedwater type events. For a reactor high pressure trir setpoint of 2300 psig, it was shown in Reference 1 that the PORV would not lift with a setpoint of

>2400 psig. The margin to the PORV setpoint can be increased, however, by use of ARTS.

Sensitivity studies on time to reach the PORV setpoint vs. power level for i a loss of feedwater event have been performed. Table 15-1 and Figure 15-1 i displays the results of these analyses. The results are far a trip on high I RC pressure since that gives the shortest time to steam generator dryout assuming no auxiliary feedwater. For potter levels <255 FP, it can be seen that sufficient time for operator action exists to initiate trip at any bypass setpoint below this valae should be a matter of providing '

sufficient operational flexibility.

For the turbine trip at low power event, the system has sufficient responsiveness such that, at lower power levels (<20%),a high pressure reacter trip is not anticipated if the turbine trips.

Stear J nerator (SG) water level is not used as an ARTS i'iput signal since as deunstrated in reference 2 (attached) it is not anticipatory of reactor trip on high pressure at high power levels. At low power levels >10 minutes is availabic for the operator to trip the reactor from the cont rol room.

Based on this SG water level is not a necessary input for the ARTS.

i 1494 015 Am. :

RE FE'HiNCES :

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I llGh' Report to the ?;RC, ?Lty 7 1979, a uation of Transient Dehavior i and Small Reactor Coolant System 13reaks }in the177 Fuel .issemb]y Plant".

2. " Anticipatory Trip Functions for 177 F.\ Plants", Document 86-1102525-00. ,

i 1494 016 dm. 7

2.1.1.5 Containment Isolation Modifications 2.1.1.5.1 System Description The functional requirements of the additional containment isola-tion signals are the following:

1. Provide diverse containment isolation signal from the appli-cable reactor trip, high radiation, 1600 psig SFAS, or pipe break signal. These signals will assure that radioactive material is not transferred out of the reactor building before a 4 psig isolation signal is reached.

2. All lines open to the containment atmosphere or connected directly to the RCS (either normally or intermittently which can result in transfer of radioactivity outside containment),

whicli are neither part of the Emergency Core Cooling Systems nor support for RCP operation, should be isolated on reactor trip.

3. In order to maintain non-ECCS support services for RCP opera-tion, the following service lines should be classified as Seismic Category I and closed on the following signals, provided that the piping is protected from pipe whip and/or jet impingement (see Fig. 2.1-5), Deletion of 4 psig RB Isolation Signal Logic):

a. Reactor coolant pump seal return valves MU-V25&26, should be isolated on 30 psig reactor building pressure signal or by the operator through remote manual operation on high radiation alarm.

b. Nucle.ir Services Closed Cooling (NSCC) water and Interme-diate Closed Cooling (ICC) water, valvee IC-V2, 3, 4, &

6, should be isolated in accordance with the logic of Figure 2.2.

c. Normal fan cooler coils will be isolated by 4 psig reactor building pressure signal and 1600 psig SFAS.

Emergency cooling will be initiated by the 1600 psig signal.

In order to utilize specific systems which have been auto-matically isolated, an isolation signal override capability is required. The isolation signal override shall be either on a total basis or on an individual penetration basis dependent on the isolation signal source and the penetration which is to be opened. The override will be to the isolation signal which will not automatically reopen the isolation valves. Operator action to reopen selected containment isolation valves will be required after the signal override has been accomplished. See Table 2.3-1 for a listing of penetrations and the required isolation override require-ments.

1494 017 2.1-11 Am. 7

2.1.1.8 Leak Reduction Program for Systems Outside Cor.cainment A leakage reduction program is being developed consistent with the requirements of NUREG-0578. Babcock & Wilcox (B&W) has been contracted to provide assistance in developing and accomplishing the overall program.

Metropolitan Edison Company, assited by BLW will implem _nt the leakage reduction program in thret distinct phases.

In phase I, the scope, plan and deve1coment will be accomplished. This will include:

1. Determination of which systems need to be included in the program and which systems may be excluded.

2. Determination of where the leakage should be measured on each system.

3. Determination of the best method to measure leakage.

4. Determination of the system and plant' conditions during the leakage measur ment.

5. Development of a testing procedure for each system.

6. Development of a method to collect and present the data such that meaningful recommendations can be made.

7. Develop a schedule and frequency for data collection to improve the consistency of sample results.

In phase 2, the actual lerkage measurement tests will be performed for those systems identified. And in phase 3, the data collected during the tests will be evaluated and the necessary corrective actions performed. The results of these phase 2 tests will be reported to the NRC within 60 days of completion of phase 3.

Phase 1 is expected to commence the week of December 3, 1979 and is estimated to last approximately 3 weeks. Phase 2 will commence immediately after completion of phase 1 dependent upon system availability. Total duration is expected to be approximately three weeks. Phase 3 will commence immediately after completion of phase 2 with an expected duration of four to six weeks.

After the program's initial implementation, Metropolitan Edison Company will initiate a Preventive Maintenance Program which will perform periodic leak tests of the systems defined in the initial program. The frequency of these tests will not exceed refueling cycle intervals.

1494 018 2.1-29a Am. 7

Systems to be leak tested will include but may not be limited to:

1. Makeup and Purification (including RCS letdown)

2. Decay Heat Removal System

3. Waste Gas System Consideration will be given to the type of potential leak paths ident.ified by NRC dated October 17, 1979 concerning the North Anna event.

A summary description of the leake;_ reduction program will be provided by January 1, 1980 and the program shall be implemented before restart.

1494 019 2.1-29b Am. 7

2.1.2.7 Increased Range of Radiation Monitors (2.1.8.b) 2.1.2.7.1 The existing Radiation Monitoring System provides in-line monitoring capability for effluents from:

a) Auxiliary and Fuel Handling Building (RM-A8) b) React ' ilding Purge (RM-A9) c) Condenser vZf-Gas (RM-AS)

Discharge from Waste Gas Decay Tanks are monitored by RM-A7 prior to combination with other exhaust and af ter dilution by RM-A8. The Reactor Building Hydrogen Purge System discharge f s monitored by the normal purge system monitor RM-A9.

The monitors, RM-A8 and RM-A9 are manufactured by Victoreen, Inc. and consist of :

a) A fixed filter particulate monitor; Beta scintillation deiector; sensitivity approxnately 1.5 x 1010 cpm / min pC1/cc based on SR-90; full range 1 x 106 cpm.

b) A Fixed Charcoal Filter Iodine Monitor; NaI detector 9 cpm ghn fixed window; Sensitivity approximate 3y 1.3 x 10 U "" I full range 1 x 106 cpm.

c) A gross gaseous monitor; Bata Scintillation detector; Sensitivity approximately 4 x 10 7 cpm 1 x 106 cpm. pC1/ce; full range d) Air sampling pump with normal sample flew of approximately 1 cubic foot per minute.

Radiation monitor RM-A5 has only a gross gaseous monitor (c above) situated on the discharge of the condenser vacuum pumps, exhausting to the suction of the vacuum pumps. Flow through the monitor is regulated to m.Antain approximately 500 cc/ min. All monitors have Control Ro a readout and recording.

2.1.2.7.2 Increased range capabilities will be provided for each of the final effluent monitors described above (RM-A8, RM-A9, RM-A5).

The additional monitoring range will be provided for each monitor by a lead shield encased GM tube located on the existing sampling lines with a collimator to increase the range of the instrument. The sensitivity of the individual inits will be determined by standard volume source calculations and verified by laboratory test. Collimators will be sized to insure the maximum range will be at least 105 h at each point. The sensitivity will assure that release' rates of:

a) 5,600,000 Ci/sec from Auxiliary and Fuel Handling Building b) 2,300,000 Ci/sec from Reactor Building Purge c) 1400 Ci/sec from Condenser Off-Gas based on maximum flow rates from each release path.

1494 020 2.1-38 Am. 7

The installation of each moniror will include evaluation cf the position of the monitor relative to other potential radiation sources and shielding necessary to minimize the effect of sources other than the sample lines on the response of the monitor. Remote readout will be provided in areas designed for habitability during an accident.

Monitors will also be provided to determine the releases from the Main Steam Safety Valves and the Atmospheric Steam Dump Valves. This monitoring will consist of radiation detectors and callimators on one of two Main Steam Lines from each Steam Generator. The monitors will be situated upstream of both the Safety Valves and Dump Valves such that the monitors will determine the total activity from the Steam Generators.

Calculations based on feedwater flow will be made to determine actual steam release from the combination of Safety and Dump Valve release points. Installation and remote readout concerns as addressed above will be included. Thr range of the instrument will be such that a release rate of 10,000 C1/sec from a single Steam Generator can be determined.

For each of the monitors described, the following applies:

a) Each will be powered from vital power, thereby providing redundancy in power supply.

b) Established sensitivities will be cor'rclated to solid source calibrations. Procedures defining calibration method will be written to assure .oper response of the instruments.

c) Emergency procedures s 11 be written to the use of the radiation instrumentaiton in conjunction with flow infor-mation to determine release rate.

1494 021 2.1-39 Am. 7

imirediate protective actions are not automatically required, declaration of a Lite Emergency will set into motion all personnel onsite and offsite that would be required to perform actions up to and including the evacuation of near-eite areas. All monitoring teams required to make continuing assessments for providing officials with information to decide on protective actions will be dispatched. The Site Emergency class includes accidents which have a signifi-(int radiation release potential. Details of all of the emergency measures that will be taken upon declaration of a Site Emergency are presented in Section 4.6.0 of this Plan.

The emergency action levels that shall require a Site Emergency to be declared shall include (but are not necessarily limited to) the following:

1. Reactor Building pressure > 30 psig.

2. Reactor coolant activity > 300 uci/ml.

3. Primary to secondary leakage > 50 gpm.

4. Loss of all offsite power and loss of both Diesel Gener-ators coincident with total loss of vital AC and DC power.

5. River stage > 307 ft. at the Rive.r Water Intake Structure.

6. Any earthquake of magnitude > SSE levels.

7. Failure of actuated Emergency Core Cooling System components to start and run follow *ng an automatic system initiation such that the number of components available is below the minimum assumed for accident analysis.

8. A valid count rate on any gaseous plant effluent monitor that would result la projected dose rate at the exclusion area bocudary of > 50 mR/hr (gamma) using adverse meteorology.

9. A valid count rate on any plant iodine effluant monitor that would result in a projected child thyroid dose at the exclusion area boundary of > 250 mR in one hour using adverse meteorology.

10. A valid dose rate on the Reactor Building high range monitor that would result in a projected child thyroid dose at the exclusion area boundary of > 250 mR in one hour or a dose rate > 50 mR/hr (gamma) using adverse meteorology and building design lenkrate.

11. Incore temperature > 700 F as measured by any two incore thermocouple readings following a reactor trip.

12. Offsite radiological monitoring reports of > 50 mR/hr (gamma) at any location.

'Ihis etaergency class, as is the case with all classes in this Plan, can also arise from an action statement in a s

Revision 1 4-6 November 1979 jfg ,

4-a . - w .- . w - . . _ . .

The major change organizctionally is that this is now a separate organization section.

(3) Modificaticas/ Operations Section - This section consists of two major sub groups, Quality Control and Operation Quality Assurance.

Quality Control is resgonsible for receiving inspection and the inspection and/or surveillance activities related to corrective maintenance, modifications, installation or new construction.

The group has specialists who are qualified to the appropriate levels of ANSI N45.2.6 and SNT-TC-1 A. Additionally, the group has a welding engineering section which reviews contractors procedures and surveils control of special processes.

Operational Quality Assurance is responsible for monitoring functional testing and performing surveillance of all operations activities. The latter includes monitoring and surveillance of plant operations, preventative maintenance, radiation protection and the processing, packaging and shipping of contaminated products, and radioactive wastes.

The Operational Quality Assurance group is also responsible for in-service inspection and monitoring performance and results of pump and valve testing to the applicable requirements of ASME Section XI.

The major difference from the previous organization is in the formation of the separate operational QA group for surveilling compliance with the technical specification requirements. The other revision is the inclusion of in-service inspection require-ments under QA Department scope.

(4) Methods, Operations and Audit Section - This section is responsible for QA Department program development. It is, therefore, responsible for coordinating activities associated with department procedures and indoctrination and training.

Additionally, the group conducts independent evaluation and assessment of the program's implementation thru Quality Assuran:e Audit Prograa.

The latter includes an evaluation of effectiveness of the pro-grammatic aspects of the QA Program. This program satisfied the requirements of ANSI N45.2.12 and utilizes auditors qualified to ANSI N45.2.23.

Assisting in this assessment is a full-time site audit group reporting independently to the Manager of Quality Assurance and the Director-Engineering Reliability through the Section Head thus providing management assessment of the effectiveness of the pro-gram. Additionally, both ections are available to provide timely close out and verification of identified problems.

1494 023 5-34 Am. 7

(5) Materials Technology Section - This is an off-site section which has the responsibility of supporting design in establish-ment and/or review of requirements. Additionally the group is available as a staff group to support Manufacturing, Construction and Operations in assessment and/or evaluation of identified materials technology problems. To help affect the implementation of this responsibility are the services of the off-site laboratory l which now reports to the Director-Reliability Engineering.

The specific services provided by the Materials Technology section include:

Non-destructive Examination In-Service Inspection Materials Engineering Welding Engineering Whereas, other sections have full-time technical expertise in these areas, this centralized gorup will provide technical direction.

The Materials Technology Group did not exist in the Met-Ed Organization although elements of materials technology function existed throughout the Met-Ed/GPUSC Organizations.

5.4.3 Program The QA Program is in the process of undergoing major revisions and ar previously identified to the NRC on October 8,1979 (GQL- 233), this revision is scheduled for submittal on .Tanuary 15, 1980, The revision was necessitated for the following reasons:

a. To describe the newly formed TMI Generation Group Organiza-tion and identify the management controls and Quality Assurance Program for this Organization. Many of the or-ganizational changes depicted are reflections of careful analysis resulting from the TMI-2 accident and our desire to provide a more effective and responsible management control over operations having an affect on nuclear safety.

b. To provide a vorkable QA program plan that describes the functional manner in which activities affecting safety are managed. Included but not limited to those activities are start-up, shutdown, normal operation, emergency actions, surveillance, maintenance, repairs, radwaste processing, and modifications.

1494 024 5-35 Am. 7

c. To upgrade the committment to Legulatory Guides and ANSI Standards previously provided under the current revision of the Operations QA Plan. This list is currently under evaluation and will be finalized in the near future and submitted as a revision to this section and the Operations QA Program.

The extent of QA coverage for each item, component, system is to be separately evaluated and graded with due considera-tion given for the following:

a) the importance of a malfunction or failure of the item to safety b) the design and fabricaticn complexity or uniqueness of the item c) the need for special controls and surveillance over process and equipment d) the degree to which functional compliance can be demon-strated by inspection or test, and e) the quality history and degree of standarization of :he item.

The TMI Generation Group is committed to a comprehensive quality assurance program consisting of a three level approach to assure satisfactory and complete implementation of the program commen-surate with its requirements for safety and performance. The program's foremost considerations are the protection of the general public's health and safety. The three level approach is defined below:

Level 1 - activities at this level include independent inspections, checks and tests. This level of activity may be performed by the Operations Department such as surveillance tests, calibration of instruments, radiation surveys, analy-ses of samples, etc., the Quality Control Section such as receipt inspection or inspections of modification or correc-tive maintenance activities, or by contractors as part of their scope of work. In all cases, the activity is performed by individuals knowledgeable of the activity being performed and qualified to perform the work. Checklists or data sheets are also used for documenting the results of activity and for providing a permanent plant record of the performance of the activity.

In all cases where the first level activities involve inspec-tien for purposes of acceptance and/or verification of modi-fications to safety systems, the activity will be performed by personnel who are independent of those performing the work.

Level T.I - The activities at this level are primarily those of surveillance or monitoring and are performed as deemed 1494 025 5-36 Am. 7

necessary by the Operational QA, Quality Control, or Manuf ac-turing Assurance departments. The level of surveillance /

monitoring applied is consistent with the importance of the item to safety. For act.ivities, whereby QC is performing first level inspectior, no second level act.vity will be required.

At this level procedures and instructions are established and surveillance records will be completed and maintained. Such surveillance / monitoring normally includes observation of quality control tests and inspections, observation of signi-ficant operations, review of records, verifications of test reports, and direct over inspection on a spot check basis.

The organizations performing this activity have these levels of authority, the lines of internal and external communica-tion for management direction and the properly trained personnel for implementation of these activities.

Level III - The purpose of this level of activity is to assure through a comprehensive program of review and auditing that the first and second levels of the program are properly functioning. The purpose of the program is also to establish that all other organizations including Operations, Mainte-nance, Engineering, Materials Management, etc. are properly satisfying all the requirements of the Operational QA Program.

At this level procedures and instructions are established including the use of comprehensive checki 3ts for documen-tation of the audit or third level activity. The program requirements of ANSI N45.2.12 are saticfied. Qualified audit personnel are included that satisfy the requirements of ANSI N45.2.23. Additional technical experts from areas with administrative reporting outside the function that is being audited will be included as the Audit Team Leader deems necessary. The organizatior performing this activity has suf ficient authority and lines of internal and external communications for obtaininr the necessary management direc-tion. In addition to the Quality Assurance Department's actiities associated with verification of the completeness and adequacy of the work performed there also are several icdependent review groups whose responsibilities are to provide independent safety review and operational advice.

These groups each have the technical expertise in the areas of licensing, operations, and radiation protection. The groups include the following: General Office Review Board (GORB) - This group is an advisory group to the Prasident of GPU and whose responsibilities are to foresee potentially significant nuclear and safety problems and to recommend to the President hcw they may be avoided.

Generation Review Committee (GRC) - This group is assigned the responsibility to provide indapendent review of desig-nated areas of operations, nuclear engineering, chemistry and radiochemistry, metallurgy, instrumentation and control, 5-37 1494 026 Am. 7

radiological safety, mechanical and electrical engineering and quality assurance practices. The GRC reviews safety evaluations, significant operating abnormalitics, viola-tions of codes, Tech Specs, etc. and deficiencies.

A member of the QA Department sits on this committee.

Plant Operations Review Committee (PORC) - This committee functions as an advisory group to the Unit Superintendent on all matters related to nuclear sefety.

A member of the QA Department - Operational QA section sits on this committee.

Radwaste Review Ccamittee (RRC) - This committee functions to provide independent review of designated activities in the areas of radioactive waste management, chemistry and radio- l chemistry, metallurgy, instrumentation and control, radio-logical safety, mechanical and electrical engineering and quality assurance practices associated with radioactive waste materials control and management.

The scope of the quality program while defined in detail in TNI procedures in general includes materials, components, systems, and structures identified by Engineering as falling under the following areas of concern:

a. Nuclear safety related

b. Fire safety related

c. Rad / Waste as required by RG 1.143

d. Rad / Waste shipments as required by 10CFR70 Appendix E

e. Quality Control augmented For new designs, QA scope coverage is expanded to include items which are "important to safety" as listed in Appendix A of 10 CFR 50. Again a graded approach will be used to detenmine the extent of QA coverage that is meaningful to apply.

As always, classifications of systems, ccmponents, and struc-tures is an Engineering responsibility subject to audit. Es-tablishment of the programmatic and inspection requirements is also an Engineering responsibility with Input and review by Quality Assurance Department - Design and Procurement Section.

QA surveillance coverage for Operations will include all items within the secondary system which are Lnportant to safety as well as present scope items.

The extent to which the program is applied is detailed in the Operational QA Program.

The Operational QA Program includes provisions for controlling, and identifying the in process, final inspection, examination, t494 027 5-38 Am. 7

test and operating status of all equipment, structures, ccmpo-nents and systems throughout the plant. This control rill include provisions which preclude the final use of structures, systems, and components which are non-conforming or removed from service for purposes of maintenance, modification, inspection or te st .

5.4.4 Procedures A documented QA review and concurrence of select procedures af-fecting Operations, surveillance, in service inspection, miin-tenance, mcdifications, engineering, etc. to assure that they have been prepared, reviewed and approved in accordance with established policy and program controls; they contain the neces-sary policy and program requirements including the inspection and verification requirements; and they contain clear descriptions relative to the extent of documenting results of completed actions.

A " graded" approach will be used as to which procedures are reviewed, so as not to involve QA in review of procedures not having a significant impact or affect on items which are important to safety. Separate or redundant QA reviews are not required or encouraged.

Procedures, instructions, drawings, etc. will be controlled to the point of use to the extent deemed necessary with due con-sideration for their impact on their importance to safety.

To assure adequate compliance to requirements, these activities as others affecting quality will be programmatically surveilled and/or audited.

5.5 STATION ORGANIZATION UNDER ACCIDENT CONDITIONS The TMI-1 Emergency Plan (See Section 4.5.1.3) incorporates all the current requirements in the area of emergency planning. It is anticipated that it will be revised once comments from the NRC Task Force on Emergency Planning are received.

The Emergency Plan is to be implemented 60 days prior to the restart of TMI-1. Upon final plan approval, specific personnel assignments will be made to fill the emergency organization positions.

It should be noted that the Emergency Plan provides for manning tca on site emergency organization within one hour and the off site emergency organization in a two to four hour time frame.

1494 028 5-39 Am. 7

TMI GENERATION GROUP Senior V.P.

V.P. 6 Director TMI-1 Director ""g"g"_Ef Director Director Director Environment, ladiological Technical TMI-2 Recovery Reliability Health & Safety Controls Functions Engineering 1494 029 FIGURE 5.4-1 Am. 7

c. discharged to the Unit 2 Evaporator Condensate Test Tanks directly, or via the Evaporator Condensate Demineralizers.

Inadvertent discharge to Unit I components is virtually precluded by extensive isolation:

a. ALC-V169 is locked closed preventing discharges to Unit 1.

b. WDL-V421 and WDL-V422 are locked closed to prevent discharges to individual components in Unit 1.

c. A pancake fitting is also installed at a spool piece in the transfer line to prevent discharges to Unit 1.

d. An existing line which was tapped to get the processed water through the yard has a tee in Unit 1. A blind flange is on the unused branch of this tee.

7.2.2 Fuel Handling Building Environmental Barrier Due to the physical iayout of the Units there is a common air space connecting tae individual units' Fuel Handling Building.

The common air s pace extends to the primary personnel access from the TMI-1 tadiation protection control point to the TMI-1 Auxiliary building. The original conceptual design for the separation of the units required the installation of a physical barrier wall that would extend to the ceiling of the fuel handling building. The wall would constitute a barrier to unrestricted personnel access and communication of the ventila-tion systems of the individual units.

Upon close examination during the engineering of the barrier, it was detenmined that the concept was impractical because of the proximity to safety related equipment, (2) limitations to the use of the fuel building crane needed for fuel receipt and (3) periodic temporary breeches in the wall during transfer of the crane from unit to unit.

An alternate concept that has been selected involves a physical isolation of the Unit 1 Auxiliary building and fuel building access way at the 305' elevation from the remainder of the fuel handling building. The air space of the fuel handling building operating levels would be. common with ventilation nndifications made to minimize the communication of air between the units.

For details of the modifications see Supplement 1, part 2

• response to question 52.

An approved environmental barrier system will be functional in the fuel handling building prior to start up of TMI-1.

7.2.3 Liquid Radwastes and Miscellaneous Waste Evaporator The original design of the TMI-1 liquid radwaste treatment system will in no way be impaired by the separation of the units. The TMI-l radwaste system was designed to conservatively 7-3 1494 030 ^=- 7

handle the waste generated from a single operating unit and TMI-l operated successfully in the single unit mode in excess of three years during TMI-2 construction. The TMI-2 design utilized the conservation of the TMI-l system to provide radwaste treatment capabilities for liquid radwaste generated from its operation.

Therefore, all TMI-2 liquid radwaste had to be transferred to TMI-l for treatment. Therefore isolation of the units will increase the TMI-l liquid redwaste capability relative to that available during the pre-accident period. The Miscellaneous Waste Evaporation is the primary process equipment used to treat liquid radwaste.

The evaporator has been used successfully in the past for a single unit waste treatment requirements. Because of interferences in TMI-l caused by the storage of TMI-5 waste waters, the evapora-tion is not currently in use. At the time of startup of TMI-l (or within the month previous to starting) the evaporation will be shown to be operational or replacement equipment shall be available to supplement evaporation operation.

Although, TMI-l radwaste processing capabilities are currently limited by the presence of TMI-2 liquid radwaste stored in TMI-l process vessels, this limitation will be completely removed by transferring and flushing all TMI-2 waste stored in TMI-l to TMI-2 facilities prior to startup. TMI-2 miscellaneous liquid radwaste will be processed in systems installed (or to be installed) in accordance with the recovery effort. The systems include Epicor 2, the submerged demineralizer system and an evaporator / solidification system.

A complete description of the TMI-l radwaste treatment system capability is provided in section 7.3.1.1.1 of this report.

7.2.4 Solid Waste Disposal Prior to the accident, TMI-2 radwaste system was completely dependent on TMI-1 facilities for solid waste processing and disposal of evaporation concentrates, spent resin and compactible trash.

Isolation of the two units will enhance the ability of TMI to process solid radwaste. TMI-2 had installed separate trash compact-ing equipment and will obtain the equipment to solidify liquid and wet solid waste as part of the recovery program. TMI-l capabilities are not required for the TMI-2 recovery program. TMI-1 currently has an opeating trash compactor used exclusively for TMI-l trash.

Solidification facilities will be available through the use of a contractor for the short term and subsequently through the use of a permanently installed system. Solidification capabilities are discussed more thoroughly in Supplement 1, part 2 response to question 53.

7.2.5 Sanitary Facility Drains By design, sewage and sanitary drains from TMI-2 would join those from TMI-l at the sewage pumping station and then be treated in a common sewage treatment plant. Operationally, the sewage from the individual unit currently remain separate 1494 031 7-4 Am. 7

APPENDIX 7A 1494 032

RPP-Unit I 11/28/79 Revision 0 RADIATION PROTECTION PLAN THREE MILE ISLAND NUCLEAR STATION UNIT I 1494 033 e

RPP-Unit I 11/28/79 Revision 0 1.0 RADIATION PROTECTION PLAN 1.1 It is the policy of the Metropolitan Edison Company to maintain the radiation exposure of station personnel, contractors, visitors, and off-site personnel within the limits of 10CFR20 and Appendix I of 10CFR50, and As Low As Reasonably Achievable (ALARA).

1.2 This Radiation Protection Plan prescribes the standards for implementing this policy. The Radiation Protection Plan and the supporting Station Health Physics Procedures form the basis for compliance with prescribed guides and limits.

1.3 All work involving radiological consideration shall be conducted according to written procedure. Written procedures regarding the following aspects of the Radiation Protection Plan shall be provided fcr out not limited to:

1.3.1 Individual medical evaluation, bioassay analysis, and monitoring of personnel exposure.

1.3.2 The evaluation of hazards associated with all work involving radio-active materials or exposure to ionizing radiation.

1.3.3 Appropriate protective measures to maintain exposure ALARA.

1.3.4 Training and indoctrination of all personnel in the policy and procedures applicable to their work in accordance with 10CFR19.

1.3.5 Calibration and operation of monitoring and laboratory devices.

1.3.6 Administration of the respiratory protection program.

1.3.7 Applicable controls of radioactive materials.

1.3.8 Auditing and surveillance of the Radiation Protection Prcgram. .

1.4 Procedures shall provide adequate guidance and specify appropriate methods or techniques to ensure that the performance of such activities are in accordance with sound radiation protection principles, and are in com-pliance with applicable regulatory provisione. Station Health Physics 1494 034 7A-1 Am. 7

RPP-Unit I 11/28/79 Revision 0 Procedures shall be prepared, reviewed, approved, and controlled as described in Station Administrative Procedures. Routine and non-routine maintenance / operations activities performed in radiological areas, subject to review by radiation protection personnel, are addressed in procedures or in applicable radiation work permits.

1.5 The effectiveness of the Radiation Protection Program is maintained by the implementation of a Surveillance Program. The Surveillance Program includes periodic quality assurance audits and periodic outside technical reviews. These methods are in addition to the surveillance conducted by line supervision and internal audits.

1.6 Quality Assurance Audits will be conducted on safety-related aspects of the Radiation Protection Program. The following aspects are identified as safety-related in Append.ix A of Regulatory Guide 1.33, " Quality Assurance Program Requirements (0peration)".

1.6.1 Access control to rafiation areas, including Radiation Work Permit system.

1.6.2 Radiation surveys.

1.6.3 Airborne raoioactivity monitoring.

1.6.4 Contamination control.

1.6.5 Respiratory protection.

1.6.6 Training in radiation protection.

1.6.7 Personnel monitoring.

1.6.8 Bioassay program. 1494 035 1.6.9 Implementation of ALARA.

7A-2 Am. 7

RPP-Unit I 11/28/79 Revision 0 2.0 RADIATION PR'TECTION ORGANIZATION 2.1 General The Radiological Controls department, within the TMI-1 organization has the responsibility for Radiological Control program design, support and enforcement.

Design Program design includes the preparation and review of the TMI-l Radiation Protection Plan and implementing procedures, Radiological Control foreman and technician training programs, and the ALARA, Bioassay and Respiratory Protection programs and implementing procedures.

Support Program support functions include the development of information related to radiclogical conditions to enable the production department to comply

,,E.h regulations and correct dork practices. Support functions also include the accomplishment of radiological surveys and the associated conditions assessment with related recomendations for protective controls and work practices.

Support also includes the monitoring of ongoing production department work for compliance with regulations and correct radiological control work practices.

Enforcement Enforcement includes both the responsibility and authority to immediately stop any and all work not being accomplished in accordance with radio-logical control practices, procedures, and regulations.

Production Department The production department under the Vice President Metropolitan Edison Company has the direct responsibility for compliance with regulations and radiological control proadures.

1494 036 7A-3 Am. 7

RPP-Unit I 11/28/79 Revision 0 2.2 Organization, Function, Responsibility, and Qualifications.

2.2.1 Function

The Manager of Radiation Protection reports tc the Senior Vice President Met-Ed on all matters concerning Radiation Protection and Health Physics associated with operations and maintenance of Unit I. In the execution of his management responsibility in the above, he directs the management staff shown in Figure 2-1,

2.2.2 Qualifications

The Manager - Radiation Protection shall as a minimum possess a Bachelcr's Degree in Physics, Engineering or a related science. He must have a minimum of 3 years experience in the nuclear power industry.

2.3 Supervisor of Radiological Engineering

2.3.1 Function

The Supervisor of Radiological Engineering reports to the Manager of Radiation Protection on all matters concerning:

. Radiation Protection Program Design

. Respiratory Protection Program Design and Review

. Effluent Assessment

. Dosimetry

. ALARA Program Design

. Plant Operations Review Committee

. NRC Interface

. Plant Modification / Procedure Modification Review

. Special Proiects ,

in the operation and maintenance of Unit 1. He directs his technical staff of Radiological Engineers.

2.3.2 Qualifications

The Sucervisor of Radiological Engineering shall be qualified as per ANSI /ANS 3.1-1978 and shall possess the qualifications 1494 037 7A-4 Am. 7

RPP-Unit I 11/28/79 Revision 0 prescribed in Regulatory Guide 1.8. In the event that the Incumbent does not meet the latter qualifications, a Deputy so qualified shall be assigneu.

2.4 Radiological Engineers Radiologic _ Engineers aport to the Supervisor - Radiological Engineering and, as av of:ci, perfo. tasks and are assigned responsibilities associated with the functions listed in paragraph 2.3.1.

2.5 Supervisor of Radiation Controls

2.5.1 Function

The Supervisor of Radiological Controls reports to the Manager of Radiation Protection on all matters concerning:

. Program Enforcement (including Respiratory Protection)

. Surveys

. RWP's

. Radioactive Material Controls

. Ins trumentation

. Direct Support of Maintenance and Operations Activities.

in the operation and maintenance of Unit 1. He directs his technical staff of Radiological Controls Foremen assigned.

2.5.2 Qualifications

The Supervisor Radiological Controls shall as a minimum have 8 years experience in nuclear power plant radiological controls at least 5 of which must be in a responsible position (Technician level or equivalent). He must have a minimum of two years experience in the direct supervision of radiological controls personnel.

The Radiological Controls Foremen direct the activities of the Radiological Controls Technicians in the fulfillment of their responsibilities.

4 2.6 Radiological Controls

2.6.1 Function

The Radiological Controls Foremen report to the Supervisor 7A-5 Am. 7

RPP-Unit I 11/28/79 Revision 0 of Radiation Protection on assigned areas of responsibility to implement and enforce the Radiation Protection Program areas as indicated in paragraph 2.4.1.

2.7 Radiological Controls Technicians

2.7.1 Function

The Radiological Controls Technicians report to their assigned Radiological Controls Foreman on assigned tasks within his groups area of responsibility.

2.7.2 Qu lifications: Technicians in responsible positions shall meet or exceed the qualifications of ANSI /ANS-3.1-1978 paragraph 4.5.2. All Technicians will be qualified via training and examination on each area or specific task related to their radiological control function prior to the performance of those tasks.

1494 039 7A-6 Am. 7

RPP-Unit I

' 11/28/79 Sr. Vice Pres Revision 0 Met-Ed T

Manager @

Health Physics N

c==2

~' @=>

c=

Supt -

Administrative Supervisor M Radiological S;. eering Assistant Radiological Control g::2 RadiologicalIr. eers Clerical i Radiological Controls Personnel Foremen Radiological Ovntrols Technicians 4

4 Figure 2-1 4

g Unit I Radiation Protection Organization ,

4 O

7A-7 Am. 7

RPP-Unit :

11/28/79 Revision 0 3.0 STANDARDS FOR RADIATION EXPOSURE CONTROL 3.1 The policy of Metropolitan Edison and General Public Utilities is that all radiation exposure shall be ALARA and within the limits specified in 10CFR20.101, 20.104, and 20.105(b) for occupationally exposed individuals and Appendix I of 10CFR50 for members of the public.

3.2 The maximum permissiSle occupational radiation exposure for individuals 18 jears of age or older will be limited to that which is specified in 10CFR20.101.

3.3 Metropolitan Edison Company shall not possess, use or transfer licensed material in such a manner as to cause any individual within a restricted area, who is under 18 years of age, to receive in any period of one calendar quarter from radioactive material and other sources of radiation within the possession of the Metropolitan Edison Company, in excess of 10 percent of the limits specified in 10CFR20.101.

3.4 To insure compliance with the limits specified in 3.2, procedures will be developed to track personnel exposure and require stricter radiological controls and a higher level of authorization as individual exposures increase. Specific administrative control levels will be defined within procedures.

3.5 The standards for occupational radiation exposure to fertile females shall be ALARA and in compliance with the guidelines of Regulatory Guide 8.U 3.6 Maximum Permissible Concent.ations in Air.

3.6.1 Occupational exposure of individuals to airborne radioactive materials is limited by the requirements of 10CFR20.103. This regulation specifies maximum permissible limits on the inhalation of radioactive materials by individuals in restricted areas.

1494 041 7A-8 Am. 7

RPP-Unit I 11/28/79 Revision 0 3.6.2 Metropolitan Edison Company shall not possess, use or transfer licensed material in such a manner as to cause any individual within a restricted area, who is under 18 years of age to be exposed to airborne radioactive material possessed by the licensee in an average concentration in excess of the limits specified in Appendix B, Table II of 10CFR20.

3.6.3 Administrative Control of Internal Exposure.

In general, exposure to airborne concentrations higher than the MPC's shall be prevented or avoided, by wearing of appropriate and properly fitted respiratory protective equipment.

1494 042 7A-9 An. 7

RPP-Unit I 11/28/79 Revision 0 3.7 ALARA 3.7.1 Radiation exposure of personnel shall be maintained at a level that is As Low As Reasonably Achievable, ALARA, within the dose limits specified in 10CFR20. The As-Low-As-Reasonably-Achievable levels shall be achieved by using the guidelines presented in NRC Regulatory Guides 8.8 and 8.10. Company Management is committed to maintaining exposures ALARA. Station procedures and practices for achieving ALARA exposures shall be reviewed and audited periodically by the Radiation Protection Depart =nt.

3.7.2 The preparation of crocedures necessary to develop a strong ALARA program should include but are not limited to:

. Administration

. Methodology

. Training

. Specific job considerations

. Program assessment

. Safety evaluation 1494 043 7A-10 Am. 7

RPP-Unit I 11/28/79 Revision 0 3.8 Emergency and Accidental Exposure (Once in a lifetime dose,)

3.8.1 The planned radiation exposures during a radiological emergency shall be in keeping with the guidance setforth in NCRP Report No. 39 " Basic Radiation Protection Criteria." It is considered that an emergency dose of 25 REM to the whole body may be accepted under unusual circumstances. It should be pointed out that every reasonable effort must be made to minimize exposure, even in emergencies.

3.8.2 Over Exposure Any individual who receives radiation exposure in excess of the limits prescribed for any calendar quarter shall be handled according to applicable procedures and in accordance with 10CFR20.405.

3.9 Classification of Areas 3.9.1 Responsibility It is the responsibility of each individJa1 to obey all radiation control procedures and report to his supervisor any circumstances where there is doubt as to the correct procedure or to the safety of the operation. It is the responsibility of supervision (foreman) to assure that all work in the restricted area is performed in accordance with approved procedures. It is the responsibility of the Radiation Protection Department to evaluate radiological conditions in the restricted area and recommend precautionary measures.

3.9.2 Restricted Area A restricted area is defined as any area access to which is controlled ,

by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials. A Restricted Area is any area in which a person can receive more than 2 mrem in any hour, more than 100 mrem in any seven consecutive days or more than 500 mrem 7A ll 1494 044 w.7

RPP-Unit I 11/28/79 Revision 0 in a year. The restricted area will normally be defined by the security " protected area" ferice, but may include specific areas outside the fence to accommodate special circumstances such as the storage of radioactive material.

3.9.3 Controlled Area The controlled area shall encompass all plant areas where radiation or contamination has a higher potential to exist in the amounts above the limits set for Clean Areas. Access to the Controlled Area will be limited to those persons authorized by the Radiation Protection Department as described in training and access control procedures.

Entry to and exit from the Controlled Area for individuals shall be through designated Access Control Points only.

Controlled Areas are established in order to comply with Title 10CFR Part 20. Controlled Areas shall be surveyed and conspicuously posted with appropriate Radiation Caution eigns as required by the criteria for Radiation /High Radiation Areas, Airborne Radioactive Areas, Contamination Area.

3.9.4 Radiation Area, High Radiation Area, Airborne Radioactivity Area.

All posting of areas will be in accordance with 10CFR20.203, except as indicated in Section 3.12.

3.10 Area Contamination Limits 3.10.1 A contaminated area is any area in which removable surface radioactive contamination exceeds the following:

. Beta-Gamma (8y) 1000 DPM/100cm2 2

. Alpha (a) 100 DPM/100cm 3.11 Internal Dosimetry An internal dosimetry program shall be established for the assessment of individual intakes of radioactive material to implement the 7A-12 Am. 7 1494 045

RPP-Unit I -

11/28/79 Revision 0 requirements of 10CFR Part 20.103 and Regulatory Guide 8.9 and the guidance contained in ANSI N343-1978. Implementing procedures will include the following areas:

3.11.1 Conditions (potential exposure) under which bioassays will be required.

3.11.2 Selection of measurement techniques and quality control criteria, measurement frequency, and program participants.

3.11.3 Specification of actions to be taken based on measurement results, with action points.

3.11.4 Interpretation of measurement results.

3.12 Exemptions to Title 10CFR20.

The Radiation Protection program shall comply with the requirements of 10CFR20 with the following exception:

3.12.1 Paragraph 20.203 " Caution signs, labels and signals".

In lieu of the " control device" or " alarm signal" required by paragraph 20.203 (c) (2), each High Radiation Area (100 MREM /hr or greater) in which the intensity of radiation is less than 1000/ MREM /hr shall be barricaded and conspicuously posted as a High Radiation Area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit and an individual or a group of individuals permitted to enter such areas shall be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area.

In addition to the above procedure, each High Radiation Area in whith the intensity of radiation is greater than 1000 MREM /hr will have locked barricades to prevent unauthorized entry into these areas.

Procedures will be implemented to administratively control the issuance of keys and to maintain that control under the Supervisor -

Radiological Controls.

7A-13 1494 046 im.7

RPP-Unit I 11/28/79 Revision 0 4.0 RESPIRATORY PROTECTION PROGRAM, 4.1 Policy The implementation of the Respiratory Protection Program shall be in accordance with the requirements of appropriate NRC Regulations (10CFR20.103), Regulatory Guide 8.15 (including NUREG-0041) and the requirements of this manual.

The primary objective of the Respiratory Protection Program is ',o limit the inhalation of airborne radioactive materials by personnel. The preferred method for achieving this objective is the application of engineering controls (process containment, ventilation systems, and local exhaust equipment). Engineering controls, including process containment and ventilation systems have been built into the nuclear power stations to remove airborne radioactive materials from the work environment and thus protect the worker. When additional engineering equipment should be used; however, it is not intended that respirators be used as a substitute for effective engineering controls; because, under some circumstances, the use of respirators can lead to an increase in internal exposure; result in additional physical and/or mental stress; and increase the risk of injury, due to interference with vision, freedom of movement, and the ability to comunicate.

4.2 Administrative Requirements 4.2.1 Procedures Written procedures for implementing the requirements of this policy, ,

shall be prepared and utilized. The following phases of the Respf.atory Protection Program require written, detailed procedures.

In implementing this program NUREG-0041, Manual of Respiratory Protection Against Airborne Radioactive "'.terials Regulatory Guide 8.15 7A- 14 Am. 7

RPP-Unit I 11/28/79 Revision 0 and 10CFR20.103 reg .lations are to be used as reference documents.

1. Selection and Supervision of Personnel (NUREG-0041 Sections 7, 12,&13).

2. Training of Personnel (NUREG-0041, Section 8).

3. Methods for Ensuring Adequate Fit (NUREG-0041, Sections 7, 8,

&12).

4. Maintenance Program (NUREG-0041, Sections 9-10).

5. Administration of Respiratory Protection Program (NUREG-0041,

'ections 2,3,4,5,6,9,10&12).

4.2.2 Records An adequate record system is necessary so that the effectiveness of the Respiratory Protection Program may be evaluated periodically.

These records should be sufficiently detailed to demonstrate compliance with NRC and station procedural requirements. Personnel records should include the ability or the inability to wear respiratory-protective equipment, the types of equipment that can be worn properly, and results of body-burden analysis. Training records should include subjects covered and results of tests. Monitoring records should be kept of the results of all air samples taken to evaluate the individual's exposure, and to properly select a respiratory protective device.

Maintenance records should show out-of-service time for respirators, common failures of particular respirator types, monthly inspection and repair of emergency respirators, and complaints on respirator design.

A periodic review of respirator usage is needed to provide information for reordering canisters and other replacement parts, and to establish a replacement table for ordering respirators and/or components when needed. ]494 Q4@

7A-15 Am. 7

RPF-Unit I 11/28/79 Revision 0 5.0 PERSONNEL MONITORING 5.1 All individuals who are subject to occupational radiation exposure and while within the Three Mile Island Restricted Area are required to have in their possession personnel monitoring devices capable of measuring the dose received from external sources of ionizing radiation.

5.2 Responsibility It is the responsibility of the Radiation Protection Department to establish and maintain the personnel monitoring program consistent with the requirements of 10CFR20. It is the responsibility of the individual to wear the TLD and the self reading dosimeter in the prescribed manner and assure their safe keeping, The loss or damage of any personnel monitoring device will require the immediate notification of the Radiation Protection Department.

5.3 The official and permanent record of accumulative external dose received by individuals will be obtained principally from the interpretation of the TLD.

The direct rer.uing dosimeter will provide day to day indication of external radiation exposure.

5.4 TLD Issue 5.4.1 Any person who enters the Restricted Area under such circumstances that he receives, or is likely to receive a dose in any calendar quarter in excess of 25 per cent of the applicable value specified in 10CFR20.101 will be issued a Beta-Gama (Sy) TLD to wear at all times while in Control Area. A neutron sensitive device in addition to the Beta-Ga..ma (sy) sensitive TLD will be issued to personnel whenever a significant neutron exposure is possible.

1494 049 7A-16 Am. 7

~

RPP-Unit I 11/28/79 Revision 0 5.5 Dosimeter Issue 5.5.1 A self reading dosimeter will be worn by station personnel when entering the controlled area of the plant. Dosimeters will be read, recorded and re-zeroed regularly. Dosimeter % cords will furnish the exposure data for the administrative control of radiation exposure.

5.5.2 Procedures will define the method of issuing, wearing and returning Self-Reading Dosimeters, recording of exposures and actions to be taken for lost or off-scale dosimeters.

5.6 TLD Process 5.6.1 The TLD will normally be processed at monthly intervals. The TLD of any individual will be processed imediately when an over exposure has occurred or is suspected, when the Self-Reading Dosimeter is off-scale or lost, or as deemed necessary by Radiation Protection Personnel.

5.7 Plant Radiation Monitoring 5.7.1 Radiation levels within the plant shall be monitored by a combination ofportablesurvey, semi-portableinstruments(continuousairmonitors, hand and foot monitors, portal monitors) and fixed monitors (i.e.

remote area monitors, effluent activity monitors). These instruments shall be maintained and periodically calibrated to assure a consistent, reliable and predictable response to ior,izing radiation.

5.7.2 frocedures will define the calibration frequencies and methods and will comply with the provisions cf AP1023.

5.8 Personnel Exposure Investigations 5.8.1 Whenever a situation occurs involving the suspected or knowr, exposure of personnel to radiation in excess of permissible limits 7A-17 Am. 7

RPP-Unit I 11/28/79 Revision 0 specified in Title 10CFR Part 20, the incident shall be promptly investigated and personnel exposures evaluated. This may require special Bioassays, Whole Body Count, Radiation Surveys, Air Samples and TLD Analysis.

The report will include the following data:

1. Description of the operation.

2. Conc'tions under which the over exposure occurred.

3. flames of personnel involved together with previous information and exposure records.

4. Amount of exposure received.

5. Recommendations fcr corrective measures to prevent similar over exposures.

5.8.2 Personnel involved in a radiation incident and whose exposure is not known shall not be assigned to Control Area work until their exposures have been evaluated.

1494 051 7A-18 Am. 7

RPP-Unit I 11/28/79 Revision 0 5.9 Radiation Work Permit (RWP) 5.9.1 A Radiation Work Permit will be required for any work or entry to controlled areas that would involve or create the following:

5.9.1.1 Radiation Area 5.9.1.2 High Radiation Area 5.9.1.3 Airborne Radioactivity Area 5.9.1.4 Contaminated Area 5.9.2 Individuals entering RWP areas must be qualified in accordance with training procedure or be provided with a qualified escort.

5.9.3 Procedures for issue and control of RWP's will be established and include:

5.9.3.1 An evaluation of radiological conditions involved, specifying all radiation safety precautions, monitoring and protective clothing that will be required for each proposed task.

5.9.3.2 Authorization requirements of both the Radiation Protection Department and Unit Operations Department prior to the start of any work.

5.9.3.3 Any applicable record keeping requirements for demonstrating compliance with 100FR Parts 19.12, Instructions to Workers, and Part 20.201, Surveys. The RWP will be also used to track individual exposure and MPC-hours.

1494 052 7A-19 Am. 7

RPP-Unit I 11/28/79 Revision 0 5.10 General Rules for work in Control Area.

5.10.1 All work is to be conducted in a n3nner consistent with maintaining radiation exposure to personnel ALARA. Thus effective use of time, distance, and shielding will be incorporated.

5.10.2 TLO's and Dosimeters shall be worn at all times. Dosimeters should be read periodically when working in a radiation area or high radiation area.

5.10.3 The required items of protective clothing shall be worn by all personnel while in the Control Area as specified by the RWP.

.i.10.4 Eating and drinking are prohibited in the Control Area. Smoking shall not be permitted in Contaminated Areas, and only allowed in designated smoking areas within the Controlled Area.

5.10.5 Visitors and/or persons unfamiliar with plant rejiation safety regulations will require an escort when entering the Control Area.

5.10.6 Personnel leaving the Control Area shall monitor themselves for contamination at the Access Control Point before leaving the area.

If a person is found to be contaminated, Radiation Protection Personnel shall be notified.

5.10.7 Personnel shall notify Radiation Protection of the malfunctioning of any Radiation Protection Equipment.

5.10.8 Personnel shall notify Radiation Protection of a known or suspected contaminated step off pad.

5.11 Personnel Decontamination ,

5.11.1 The Radiation Protection Personnel will be notified in the event that anyone in the plant is found to be contaminated with radioactive materials. Applicable procedures concerning personnel decontamination are provided for:

1494 053 7A-20 Am. 7

RPP-Unit I 11/28/79 Revision G

. Decontamination process

. Release limits

. Alternative decontamination methods

. Documentation 1494 054 O

9 7A-21 Am. 7

RPP-Unit I 11/28/79 Revision 0 5.12 Removal of Material and Equipment from the Controlled Area.

5.12.1 Any component, item of equipment or tools having been used in the Control Area, will require a Beta-Gamma (sy) dose rate and smear survey prior to removal from the Controlled Area.

5.12.2 Release Limits Material and equipment will be given an unconditional release by Radiation Protection for use outside the boundary of the controlled area if detectable contamination is less than 1000 DPM/100cm2 Beta-Gamma (sy) and less than 100 DPM/100cm2 Alpha (a) contamination is present an/. raoiation levels at one inch are less than .4 Mr/hr.

1494 055 7A-22 Am. 7

RPP-Unit I 11/28/79 Revision 0 6.0 RADIOLOGICAL DROTECTION TRAINING 6.1 Radiatinn Protection Training Objectives 6.1.1 The primary objectives of the Radiation Protect. ion Training Program shall be to:

6.1.1.1 Ensure that all personnel involved are instructed about the biological effects of radiation and the risks associated with the acceptance of radiation exposure.

6.;,1.2 Provide the information needed to enable each person to comply with Health Physics procedures and respond properly to warnings and alarms under both normal and emergency conditions.

6.1.1.3 Provide the information needed to ensure that individuals can maintain their own exposure ALARA and ensure that ALARA considerations can be appropriately reflected in decisions which affect the exposure of others.

6.1.1.4 Provide the information needed to enable each person to comply with NRC regulations and TMI Unit I Technical Specifications.

6.2 Radiation Protection 6.2.1 General Radiation Protection training shall be given to personnel whose work assignments do not require unescorted access to areas controlled by Radiation Work Permits.

6.2.2 Intermediate Radiation Protection training shall be given to personnel whose assignments require unescorted access to areas controlled by Radiation Work Per lits. .

6.2.3 Initial training shall be given for the qualification of new Radiation Protection Technicians.

6.2.4 Retraining shall be given to all Radiation Protection Technicians.

Management has a strong commitment to ensure that retraining is being accomplished adequately. 1494 056 7A-23 Am. 7

RPP-Unit I 11/28/79 Revision 0 6.2.5 Respiratory Protection Training shall be given to personnel whose work assignments require that they wear respiratory protection equipment.

6.2.6 Task-specific Training shall be given to personnel who have been assigned to t task which involves significant radiological hazard.

Indoctrination in the specific hazards in the work place, in appropriate spacial radiation protection procedures, ALARA considera-tions, and the use of special protective clothing, respirators, and equipment should be considerations for the Task-specific Training.

6.3 Procedures 6.3.1 Procedures will be written to define:

. Personnel required to receive training

. Training course content

. Frequency of training

. Personnel exemptions 7.0 RADIOLOGICAL SURVEYS AND RECORDS 7.1 The Radiation Protection Program shall include radiation surveys for airborne activity, removable surface contamination and radiation levels.

Those surveys shall be performed in accordance with applicable Health Physics procedures and regulatory requirements. Surveys are performed in order to:

7.1.1 Monitor the suitability of control measures, 7.1.2 Evaluate the need for additional controls, 7.1.3 Evaluate trends for ALARA purposes, and, 7.1.4 Evaluate radiological conditions in areas routinely entered without radiation work permits coverage.

1494 057 7A-24 Am. 7

RPP-Unit I 11/28/79 Revision 0 Surveys shall be performed in

. Controlled Areas

. Uncontrolled Areas 7.2 Surveys in Controlled Areas will be performed where radiation materials may exist and appropriate radiological control measures may be implemented for the protection of personnel. Applicable procedure (s) shall be provided regarding the following aspects of these surveys:

. Type survey required

. Frequency

. Appropriate instrumentation

. Procedure

. Limitations

. Documentation 7.3 Surveys in Uncontrolled areas are provided to insure the effective controls of radioactive material. Applicable procedure (s) shall be provided regarding the following aspects of these surveys:

. Type survey required ,

. Frequency

. Appropriate instrumentation

. Procedure

. Limitations

. Documentation 7.4 Records ,

7.4.1 Records and reports required to show compliance with Federal and State regulations will be maintained on permanent file. All information from routine and special radiological surveys will be recorded on appropriate survey forms. Radiation Protection Program records w'iether considered 7A-25 Am. 7 g

RPP-Unit I 11/28/79 Revision 0 primary, secondary, or supporting records shall be maintained until otherwise authorized by the NRC and/or State authority.

8.0 CONTROL AND ACCOUNTABILITY OF RADI0 ACTIVE MATERIAL 8.1 Control and Accountability of Radioactive Material This section prescribes radioactive material control and accountability procedures to assure compliance with all applicable Federal and State Regulations regarding the transfer, possession and use of by-product and special nuclear material.

8.2 Responsibility 8.2.1 It is the responsibility of cognizant supervision to notify Radiation Protection of any impending shipment or receipt of radioactive matccial. All requisitions for radioactive materials shall be reviewed by the Radiological Controls Supervisor to ensure license possession limits are not exceeded. It shall be the responsibility of the Supervisor - Radiological Controls to monitor the shipment and receipt of all radioactive material to and from the plant.

8.2.2 It shall be the responsibility of the Supervisor - Rad Waste to coordinate and ensure with Supervisor - Radiological Controls that all shipments of Radioactive waste are monitored and comply wiLh procedures.

8.3 Procedures shall be developed for control and accountability to include:

8.3.1 Receipt, handling and storage of radioactive and special nuclear material.

8.3.2 Inventory Control

~

8.3.3 Shipment and transfer of material to other licensees. .

8.3.4 Q.sality Assurance provisions shall be incorporated into Health Physics procedures, to assure that all shipments of radioactive materials from Unit 1 are performed pursuant to 10CFR Part 71.

7A-26 1494 059 Am.7

~

RPP-Unit I 11/28/79 Revision 0 8.4 Liquid and Gaseous Radioactive Waste 8.4.1 The controlled release of all liquid and gaseous radioactive wastes 1494 060 4

7A-27 Am. 7

RPP 11/28/79 Revision 0 shall be As Low As Reasonably Achievable and comply with applicable provisions of the unit Technical Specifications and Regulatory Requirement 10CFR20.106.

8.4.2 Radioactive waste discharges shall be controlled administratively by the use of Liquid Release Permits and Gaseous Release Permits. The completion, approval and use of these permits is described in applicable Health Physics Procedures.

8.4.3 Effluent streams shall be sampled prior to release, and appropriate sampling and/or monitoring by instrumentation shall be performed during the releases, in accordance with unit Technical Specifications and appropriate Health Physics Procedures.

8.4.4 Monitoring equipment in the effluent streams shall be operable and periodically calibrated in accordance with unit Technical Specifications.

8.4.5 Systems which are not normally expected to be contaminated and which may be routes of potential contamination will be sampled on a periodic basis to avoid the possible release of these liquids in the unlikely event that contamination is present.

8.5 Solid Radioactive Waste 8.5.1 Solid radioactive wastes are processed and packaged for off-site disposal at licensed waste facilities. Solid radioactive waste includes: solidified liquids / resins, filter sludge; compressed waste-plastic / paper / damaged protective clothing, etc.; pieces of waste equipment / piping / components; and other similar materials. .

8.5.2 Radioactive wastes are prepared for solidification and solidified in accordance with operating procedures under the cognizance of a Radwaste Foreman or individual designated by the operating procedure for the earticular process.

4 gj 7A-28 Am. 7

RPP-Unit I -

11/28/79 Revision 0 8.5.3 The amount of solid radioactive wastes generated as a consequence of operating and maintaining Unit I shall be maintained ALARA.

1494 062 7A-29 Am. 7

RPP-Unit I 11/28/79 Revision 0 9.0 Investigation and Notification of Incidents .

9.1 Formal notification of various govermnental agencies shall be required in the event of a serious radiation incident or emergency as per the Emergency Plan. To assure compliance with the requirements of 10CFR20.403 and 405 all communication with the Federal and State agencies shall be handled in accordance with prescribed procedures which specify all communication requirements in connection with radiological emergencies and over exposure incidents.

9.2 All radiological emergencies and over exposure incidents shall be investagated. This investigation shall be conducted to determine the probable causes of the incident and the management systems which should be improved to prevent a recurrence.

1494 063 7A-30 Am. 7

b. negative feed to steam dif ferential pressure.

c. loss of all four reactor coolant pumps,

d. loss of both main FW pumps.

8.2.2 Modification as Result of Order of May, 1978 Modifications to the high pressure injection system. The HPI injection lines have been cross connected to assure acceptable results from a break in a high pressure injection line. Cavitat-ing venturis have been added to provide the proper flow split in the event of an HPI line break.

8.2.3 Modification Originating from within Met-Ed

1. Post accident instrument and valve operator availability vill be improved by the addition of heat chrink tubing.

2. The switchover of the ECCS system suction supply from the borated water storage tank (BWST) will be accomplished automatically rather than by operator action.

3. The reactor building spray system will be modified to delete sodium thiosulfate. Sodium hydroxide will be retained.

This change will provide equal drawdown of the BWST and NaOH tanks for a large spectrum of single failures.

8.2.4 I&E Bulletin 79-05C Met-Ed has committed to install an automatic reactor coolant pump trip to be initiated on a SFAS coincident with an indi-cation of a large (in excess of 10-20%) void fraction. (See section 2.1.2.5) 8.3 EFFECT OF CHANGES ON SAFETY ANALYSIS Following are summaries of the accidents listed in Table 8.2-1.

Table 8.2-1 indicates where FSAR analyses took credit for non-safety grade equipment, or where mitigation is dependent on a specific operating / emergency procedure or design margin. These conclusions will continue to be revised to account for plant design changes.

The event description and mitigating equipment are for the plant design before modification. The modifications discussed in the previous secitons were considered in the review of each accident.

If a modification affected that analysis, then a note as to its safety significance was made under the " conclusions" section.

8.3.1 Rod Withdrawal from Startup (FSAR Section 14.1.2.2)

1. Description Uncontrolled reactivity excursion starting from a suberitical condition of 1% Ak/k at hot standby.

8-2 1494 064 Am. /

8.3.11 Fuel Handling Accident (FSAR Section 14.2.2.1 and References 8 through 10)

1. Description Failure of a spent fuel assembly, either in the fuel handling building or inside the containment building can result in release of radioactivity to the environment. The fuel handling accident in the fuel handling building considers a 72 hr. decay period for the fuel with the release of gap activity from the entire row of fuel pins on one assembly.

100% of the noble gases and 1% of this iodine inventory is released from spent fuel pool. The fuel handling accident inside containment assumed failure of an entire assembly, filtration by the refueling canal water, and release via the purge exhaust filtration system.

2. Acceptance Criteria

i. Doses should be appropriate within the guidelines of 10CFR100 (less than 100 REM) .

3. Mitigation

i. Filtration of releases through the fuel handling building ventilation system.

ii. Filtration of releases by the purge exhaust filter system for the accident inside containment.

iii. Meteorological dispersion of 6.8 x 10-4 sec/m3 for the accident initiating inside containment.

4. Conclusion The Fuel Handling Building Ventilation system modifications af fects the method of analysis of this event. See the re-sponse to Question 52 of Supplement 1, part 2.

8.3.12 Rod Ejection Accident (FSAR Section 14.2.2.2)

1. Description Failure of a pressure barrier component could result in the rapid ejection of a control rod from the core. A power excursion and leakage of radioactive primary system fluid to the secondary side would result. Releases to the environment result both from releases via the secondary system and from leakage from containment.

2. Acceptance Criteria

i. The reactor coolant pressure boundary is not further degraded as a result of the ejected rod (no reactor vessel deformation).

8-11 Am. 7 .

Key assumptions for the small break LOCA analyses versus the TMI-1 plant design are given below:

BAW-10103 Generic TMI-l Reactor Power (MWt) 2772 2335 Reactor Trip (psig) 1900 1900 RC Pumps (LOOP) Coastdown Loastdown AFW Available** Yes-40 sec. Yes****

ESFAS HPI (psig) 1600 1600 Operator Action Yes-cross-connect none***

HPI Distribution 70% to Core 70% to core within 10 min. from time zero***

HPI Flow (gpm) 450 at 600 psig 500 at 600 psig

• • / mount assumed for generic analyses 550 gpm which is less than the minimum 900 gpm available for TMI-1. Results of Reference 2 demonstrate that EFW is not required before 20 minutes.

• • • Prior to startup TMI-l will install HPI injection leg cross connects and flow control devices to eliminate operator action to cross connect HPI and equalize flow in all four injection legs.

• • • • For worst case LOCA in which offsite power is lost, EFW is initiated by the control grade loss of feedwater signal.

In all cases, TMI-l plant specific information is as conservative or more conservative than the generic assumption.

Since the TMI-2 accident, greater focus has been placed on small break LOCA's and the capability of the ECCS to mitigate them.

Problems such as those discussed in Reference 21 (where the pressurizer stays full due to the loop seal arrangement despite loss of RCS inventory) have been addressed. These studies are documented in B&W's " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant" May 7, 1979 (Reference 2). Breaks of 0.01, 0.02, and 0.07 ft.2 are analyzed utilizing varying assumptions on the availability and timing of AFW and HPI. These analyses use the same initial assumptions as used in BAW-10103 except that ESFAS is assumed to occur at 1350 psig. Therefore, they are also bounding assumptions for TMI-1 except for the distribution of HPI flow as discussed below. The analysis in Reference 2 also established that EFW flow is not required less than 20 minutes before any steam line break accident.

8-15 Am. 7

QUESTION

5. Your response to Question 6 is not complete. Provide qualification documentation for the new "s.'ety related" control on EFW flow measuring devices.

RESPONSE ,

The qualification documentation for the EFW flow measuring devices is attached.

1494 067 Am. 7

241 flow display . . l u computer specificat. ion l r

\  ?

i PERFORMANCE \

r Zero Stability

• 015 ft/sec Linearity 1% of reading aoove 1 fps. [

Accuracy

• Quoted per application on i request f

Actual Flow Calibration:

intrinsic Calibration:

0.5-1.5% of Actual Flow " {5 240 transducer 14% of Actual Flow "

- - etiusec  ?

Data Scatter: 3 digits s.d. at zero flow Flow Range 0 to 30 f t/sec (mmimum)

]

4 Digital Display Resolution 20,000 digits (two ranges)

Temperature: [

Digital Display Update Interval 4 seconds  ?

-40* F to + 250'F: (Standard)

(No Frost or Gas Vapors in Pipe) Analog Update 'nterval 0.03 sec (intrinsic) Y Cryogenic Temperatures up to +1000*F: 1/10 sec damping included  ?

(Special Order) Totalizer Update Rate Continuous (real time) h,,

Pipe Noise / Shock: Liouid Char.v:teristics i glandard: Homogeneous.

insensitive to 120 db noise /20g shock t No undissolved solids or ^

Construction: es Lf Epoxy potted plastic and alumbium: @o t sec.

(Standard) Approval Mequired: Non-homogeneous $

Epoxy paint and stainless steel: liquids or undissolved i solids in liquid 3 (Special Order)

. ror noninai pipe size. wan inickness. in specified pipe /

240N: rnatenal. and for homogeneous non-serated liquids with Condulet Fitting: Waterproof, weather- '""## " '

• d'"""'**'" *

• y proof, olitight . Clamp mount. - ~- -

" ~ ~~~ ~ ~ ^ ~ ~ T 240P: j ENVIRONMENT j Oulck Disconnect: Weatherproof if Temperature -20* F to + 110*F connector caps are used. Spring mount. 7 Rating: Vibratior./ Shock

m-teme unn avaaavei 7 Suitable for portable and i intrinsically saf e. factog use Pipe Sizes:

g 1" to 60"; Standard. Advise rnaterial,

< + - - -

Ei diameter and wall thickness. CONSTRUCTION Below 1" and above 609 (Special Order) 241P,241R J

Lined pipe, cement pipe or open stream:

Nema 1 (Portable / It (Special Order) + 241 N Rackmount-indoor use) g Nema 4 (Outdoor or 4 Factory-wall mount) -

241E Class 1. Group D  %

(Explosion proof- 7d indoor / outdoor) 241MP 4 Gasket Sealed (Multisize i portable-indoor / outdoor)

• CC ci e POWER 50. watts 3 241P, 241 M P, 241R :15V,50/60 Hz 241N,241 E 115/230V,50/60 Hz  %

CORPORATION i We invite your 6nQuiry on special requirernents or updated specifications, d 155 PLANT AVENUE, HAUPPAUGE L.l.,

NEW YORK 11787 * (516)231 3600 2

, d 1494 06

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QUESTION

14. Your resp <nse to question 10j provided in Amendment 5 is not complete.

Provide legible arrangement drawings for the EFW system showing the location of all system pumps, piping and valves. Provide qualifi-cation documentation which as.eures that the motor driven EFW pumps will start and renain operational under the environmental conditions (humidity and temperature) resulting fram a postulated break in the main steam supply line to the turbine driven EFW pumps. Further, verify that the EFW control valves and actuators are qualified to function under these environmental conditions. Also, provide an analysis which justifies the environmental conditions (323*F) assumed as a result of the postulated steam line break.

RESPONSE

As described in response to Question 10j (Supplement 1, part 1) the subject break was not considered probatle enough to warrant detailec design consideration at the time TMI-1 was licensed. Since TMI-1 was licensed, NRC acceptance criteria for EFW systems has been modified and the EFW system has taken on new importance. In recognition of this fact, Met-Ed has initiated a complete design review of the EFW system to upgrade it to the current licensing criteria to the extent practicable. This review will consider and resolve the type of concerns raised by questions 12, 13 and 14 above. We believe that this approach is preferred over an item resolution of issues. Nevertheless, a response to your specific concerns is given below.

The EFW pumps have been certified to withstand the calculated environment.

A copy of the motor qualification certification is attached, together with the calculations which support the environmental conditions (323*F).

Environmental qualification of EF-V30A/B to 323*F was not invoked as part of the original purchase order for these valves, however, efforts are under way to determine if these valves can be certified to withstand the accident environment.

Arrangement drawings showing the location of important EFW valves and piping was provided separately on November 28, 1979 (

Reference:

E-304-086, Rev. 14).

I494 0670 Am. 7

1 .

'. ATTACHMENT TO RESPONSE 1

. f W85tingh0USD Electric Corporation

'10 QUESTION 14 0F SUPPLEMENT 1, PART 2 \* Larp AC and DC Mocer Division Box ass. Buffalo. N. Y. 4uo October 9, 1973

~~. .. Gilbert Associates, Inc.

./ Engineers and Consultants P.O. Box 1498,.6~25 X. w .d4 d W. -

Reading, Pa. 19603

Dear Mr. Shipper,

~

The emergency feed water pump motors B/M EJ-8 450 HP motors will definitely start, even if exposed to the environmental conditions mentioned in R. Broman's letter June 28, 1973. This is based on 12 PSI or less on the motor following a main steam line break.

Very truly yoi - .

0

{

YW vave Cullis .

A Senior Product Repreau *ative DC/lp . .

cc LAC Sales Manager - F. J. Wenzel .

ec: IAC Engineering - Pat Lucey 1494 071 .

e' Am. 7

.p.

\.......[

. ~

GILBEllT ASSOCI ATES, INC.

ENGINEERS AND CONSULTANTS P.O. BOX 1498 / READING. PA.19603 July I,0, 1973 GAI/THI-1/2321 Mr. C. H. Rice Westinghouse Electric Co. -

540 North 16th Street Allentown, Pa. 18102

Subject:

Three Mile Island Nuclear Station Unit No. 1 Emergency Feedwater Pump Motors B/M EJ-8 450 H.P. P.O. 86297

~

Gertlemen: t The attached letter and curves indicate the environmental conditions which will occur in the Intermediate Building following a main steam line break.

The emergency fe'dwater pump motor, located in the Intermediat.e Building, will be subj,ected to tihese environmental conditions. It is our concern, as

.- to the ability of this motor to start following subjection to these environmental conditions described in the attachment. It is our contention that the motor space heaters are to be energized at all times when the motor is not in operation.

Your immediate reply to this question would be greatly appeciated. Thank you very much for your assistance in this matter.

Very truly yours, s /

.G Y h.

P. J. Shipper, Jr.

t A L)-

P'roject Electrical Engineer PJS:psv Enclosure cc: R. W. Heward, Jr. (4)

W. T. Gunn S. Levin W. F. Sailer

!494 072 D. W. Birkhimer .

R. J. Zula R. H. Fleming J. E. Behen .

S. Hunt RE ADING OF FtCES: $25 L ANC ASTER AWENUE. TELEPHONE 21b376 3071/ CREEN MILLS. TELEPMO*st 2f k77%

, , ,', :., Ci:1* O C2'U.!CEU %.*UIldiIV'h :OI'l I. W. gYl!

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t*;euw a sr .rn, c.tur.~on';1.4 c::cs (403) 57 4 1490 June EU,1973 t GIL-O'4-11 r~, m e c'W4;;;:.,. ..Y. e~D.:

JUN 261973 .

10:. 4. I . re.ner -

W . L'. S M T il Clitcrt A:. ccintua, ic:.

5 25 L*.0: r. . c r bl .-d .

nesning, re.nr.nylv:r.ir. f -

Subject:

kvlysis of hterndiste sui 10.icc 13tirce.n=nt Follevi.m P.cin Ster.2 L-be Erer3.

'Tttrec Kile Isltnd Fuelcar Ststicu Leur l'.r. Sniler:

Attunbed r.rc thrCc crrecs shc-d.r.r. tiue historiec cf tc. peuture and ruir.tive hunidity in the inte:redir.te bui' dit;- folhviss e rnis etecr. brer}.. T:m tc';perhi.u.et O tr.* .: L';.: bre ri p'irtted Cp *.v0 tilf*s'OrCDt CEp32 E0 thSt the EC616 COUld De c).3EdO1 in the initifa. prrt,icn cf the trar.3dCDt.

At: Ind.icate:1 by the curVC",, thC DUildiu.C TEEChEC G p?32; tczparature cf 323"F 61.icut C*te c::end .:.fter the i.rtr.t. dYorn to E12"? }UTE Stein lu EL0ut Che Ph.inute, CU'1 r5turas to tahit:r.t in :f:c.ut onc hcUr. .

Vc !'IC CU!r( Stl';" prSpr.!1EC the filr.1 rejor* cn Lhe cAnlyGir., Und vill Adl it Out C0rly Dext VECh.

Vcry truly yours.

i e) i f 0 Y 0. r ${p~ L

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TABLE IS-1 P_OWER LEVEL SENSITIVITY,'

TDIE TO REACH TD1E TO FILL POWER LEVEL PORV SETPOINT PRESSURIZER 100% 3 min. 10 min. ,

75% 6 min. 11 min.

50% '12.3 min. 13 min.

25% >>15 min. 16.6 min.

e e

g .

INITIATION WHICH RESULTS IN THE SHORTEST ACTUATION TDiES. REACTOR TRIPS ON HIGH RC PRESSURE TRIP (2300 psig) .

1494 077 .

CHAtlGE IN REACTOR C00LAfiT SYSTEM .

PRESSURE VS TI!*.E TO TRIP FOR A LOSS OF MAIN FEEDWATER FROM 100% POWER

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. 1494 078

ANTICIPATORY TRIP FUNCTIONS FOR 177 FA PLANTS 86-1102525-00 Document Identification Prepared by I' D Reviewed by 25 Approved by ,r, I 6 /

f, /

1494 079

86-1102525-00 INDEX

1.0 INTRODUCTION

2.0 ASSESSMENT OF POSSIBLE ANTICIPATORY TRIPS 3.0 FUNCTIONAL ANALYSIS

4.0 CONCLUSION

S AND

SUMMARY

APPENDIX A: LOSS OF ONE FEEFPUMP ANALYSIS APPENDIX B: SAFETY EVALUATION PROGRAM FOR ANTICIPATORY TRIPS 1494 080 3

86-1102525-00 l.0 INTRODUCTION For the purposes of this report, an anticipatory trip is defined as a trip function that would sense the start of a loss of OTSG heat sink and actuate much earlier than presently installed reactor trip signals. Possible anticipatory trip signals indicative of changes in OTSG heat removal are: turbine trip, loss of main feedwater, low steam generator level, and low pressurizer level.

This report evaluates the effectiveness of anticipatory trips compared to the existing high RC pressure trip for a LOFW. Qualitative and quantitative arguments are presented which support elimination of the level trips in the pressurizer and steam generators from final desiEn considerations of anticipatory trips.

Functional response is presented in terms of a parametric study of time to trip. Thus, irrespective of the plant specific trip signals and actuation time, the hardware design can proceed with greater flexibility.

That is, by presenting system parameters, such as pressurizer fill time, as a function of time to trip, then if one plant's turbine trip signal occurs 2.1 secs af ter initiation of the event and another plant's trip signal occurs at 2.5 secs, this study will still be applicable to bath.

Some of the results presented in this report have already been sub-mitted to the NRC in Reference 1, the balance of the information will be submitted by May 21, 1979. The analyses are performed with the revised setpoints, i.e., high RC pressure trip at 2300 psig and PORV setpoint at 2450 psig. It is shown that anticipatory trips provide additional margin between the peak RC pressure af ter the reactor trip and the PORV setpoint, but provide little additional margin in the longer term re-pressurization to the PORV setpoint with continued delay of auxiliary feedwater initiation.

il

86-1102525-00 2.0 ASSESSMENT OF POSSIBLE ANTICIPATORY TRIPS In accordance with Directive 79-05B, an evaluation for design basis for anticipatory trips on turbine trip, loss of main feedwater, and low steam generator level has been completed. One of the trip functions investigated was determined not to be anticipatory as discussed below:

Low steam generator level has not been recommended as an anticipatory trip function. Figure 2-1 shows the OTSG start-up level from site data and the CADDS calculated OTSG mass inventory as functions of time following the TMI-2 event. The time of reactor trip on high RC pressure is noted on the figure and clearly demonstrates that a steam generator level trip would not have been anticipatory for a level setpoint that would not interfere with normal operations and maneuvers. The initial rapid fall in OTSG level occurs as the turbine stop valves close, momentarily stopping steam flow out of the generators.

The mass inventory increases during this period due to the loss of flow friction AP. By the time the reactor trip occurs, at 8 seconds, steam flow is re-established through the bypass system, flow friction AP re-estab-lishes the level and both mass and measured level start to decrease uniformly.

An OTSG level trip set to trip on the initial drop shown in Figure 2-1 would need to be set restrictively high for normal plant maneuvers and/or lower power levels.

1494 082

86-1102525-00 Further level information (in terms of mass inventory) is given in the figures for the analysis in Section 3.0. The results for those cases also indicate that the steam generator low levC. trip function would not be sufficiently fast to be considered anticipatory.

Anticipatory trips for loss of feedwater and turbine trip can be designed to trip the reactor in a more expedient manner than the high RC pressure trip for some overheating transients. An anticipatory trip will provide more margin to PORV setpoint during the initial overpressurization resulting from loss of feedwater and/or turbine trip. These trips will provide slightly more time to PORV setpoint and pressurizer fill for delayed auxiliary feedwater initia-tion conditions.

1494 083 A

8e-1102525-00 Figure 2-1 LOFW (TMI-2 EVENT) 2.0 -

160 TRIP ON HIGH RC 1.6 p PRESSUPE E m 120 g g

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0 0 I I I I O 20 40 60 80 Time, seconds

---

TOTAL S/G MASS, CADDS START-UP LEVEL, TMI-2 SITE DATA 1494 084

OO-AAU6J4J UU 3.0 FUNCTIONAL ANALYSIS A series of LOFW evaluations was performed at 100% full power (2772 MWt) with a reactor trip assumed on an anticipatory signal. With the new high RC pressure setpoint of 2300 psig, a reactor trip would be expected at about 8 seconds after the LOFW. The anticipatory trip study considered reactor trips with 0.4 sec,2.5 sec, and 5 sec delays from time zero. These studies also in-cluded sensitivities to AFW failure and reactor coolant pump coastdown.

The anticipatory trip study modeled a generic 177 FA plant, and is con-sidered applicable to raised or ~owered loop designs. A feedwater coastdown similar to that estimated to have occurred at the March 28th TMI-2 event was used te generate separate heat demands for each CADDS analysis. The heat demands will chcnge as the reactor trip time is delayed, because the additional heat input will boil off the fixed steam generator inventory at dif ferent rates.

For the casea where AFW flow was modeled,1000 gpm was assumed, starting at 40 seconds. With proper steam generator level and pressure control, the system parameters will begin to stabilize at 195-290 seconds, depending on trip delay time and RCP operation; see Table 3-1 and Figure 3-12. The PORV will not be actuated, nor would the pressurizer fill or empty.

With the assumption of no AFW, the PORV will be actuated about threa minutes into the event, as a result of system swell; the pressurizer fills as 10-12 minutes (see Table 3-1). A delay of reactor trip of 2-3 seconds is seen to reduce PORV time to actt. ate by about one minute, and pressurizer fill by about 2 minutes For PORV setpoints other than 2450 psig, the times will vary and can be determined from Figures 3-3, 3-4, 3-8, and 3-10.

In each of these cases, the mass addition and cooling effect of expected make-up system operation is not modeled. One make-up pump running will add about 10 inches per minute to pressurizer level, and sl/2% heat demand. It should be noted that the Mcy 7 report used a heat demand which reproduced the TMI-2 LOFW event; it has been reported by the operator that two make-up pumps were running from 13 see into the event, creating a higher heat demand than 7 1494J1R L

86-1102525-00 the anticipatory trip studies of the report assume. This is shown in Figure 3-12.

The steam generator heat demands, reactor power, RC system pressure, pressurizer level, and RC inlet / outlet temperatures are given in Figures 3-1 through 3-5 for the trip at time zero case and Figures 3-7 through 3-11 for the trip en high RC pressure (t=8 secs) case. The effects of delayed auxiliary feedwater initiation are also shown on the high RC pressure trip

Curves, 1494 086

86-1102525-00 TABLE 3-1 LOW EVENT (LOW at T=0 sec)

TIME OF REACTOR REACTOR AUXILIARY PORV PRESSURIZER S/G Lvl CONTROL TRIP (" DELAY") COOLANT PUMPS FEEDWATER OPERATES FULL (400") (P stm=1025 psig) 0.4 Run at 40 see - -

195 sec 2.5 Run at 40 see - -

225 sec 5.0 Run at 40 see - -

275 sec 0.4 Run None 235 sec 790 oec -

2.5 Run None 180 sec 685 see -

5.0 Run None 140 sec 575 see -

0.4 Coastdown at 40 see - -

255 0.4 Coastdown None 190 sec 700 see -

LOFW EVENT - TIME =0 see REACTOR TRIP AT 2300 PSIG ,

TIME OF TRIP RCP AFW PORV PRESS. FULL S/G LEVEL CONT.

8.0 Run at 40 see - -

260 sec 8.0 Run None 175 sec 620 see -

1494 087

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86-1102525-00 Figure 3-12 LOSS t,7 FEEDWATER AT T = 0 SEC NO AUXILIARY FEEDWATER l 1 i i i i 800 -

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RCP C0ASTDOWN O

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86-1102525-00

4.0 CONCLUSION

S AND

SUMMARY

A spectrum of delay times, representing anticipatory trips, has been analyzed for the loss of fadwater transient. The spectrum included trips at time zero, with a 0.4 second instrument delay, up to high RC pressure trip at time 8.0 seconds. Since a high RC pressure trip occurs very soon af ter a loss of heat sink (overpressurization) transient from 100% FP, only turbine trip and direct loss of feedwater detection trips would be considered Anticipatory.

For all tr3;a considered, including high RC pressure, the PORV is not actuated when noncal system operations occur. The pressure rise in the primary side is less for the anticipatory trips providing additional margin to PORV lift. If auxiliary feedwater is significantly delayed, then an anticipatory trip will, at best, provide about 1 minute additional time to PORV lift and about 3 minutes additional time to filling of the pressurizer. These results can be seen in Table 4-1 which shows the sequence of events for a LOFW transient with trip on high RC pressure (2300 psig) and trip at time zero.

1494 100 2.3

86-1102525-00 TABLE 4-1. LOFW-SEQUENCE OF EVENTS COMPARISON 40-s 120-s TRIP AT ZERO, EVENT AFW DELAY NO AFW NO AFW Loss of feedwater initiated 0 0 0 0 (trip occurs)

(0.4 delay)

High-pressure trip (2300 psig) 8 8 8 a PORV opens (2450 psig) a a 175 235 Peak RCS pressure 10 10 175 235 Pressurizer full a a 620 790

• Does not occur for these cases 1494 101

86-1102525-00

REFERENCE:

1) B&W Report to the NRC, May 7, 1979, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant".

1494 102 e #

86-1102525-00 APPENDIX A LOSS OF ONE FEEDWATER PUMP A special analysis was performed at the B&W Owner's Group request.

This analysis considered the loss of one main feedwater pump with the plant operating at 100% FP, RC pumps running, no power runback or auxiliary feedwater initiation and a RC high pressure trip setpoint at 2300 psig. The base parameters for this study are the same as those used in the " realistic" analysis presented in Section 4.2 of the May 7, 1979, B&W Report for 177 FA plants.

The objectives of this study were two-fold:

1) Determine if the PORV will lif t under a losa of one feedwater pump situation, and,

2) Determine if the OTSG level would be a viable anticipatory trip, i.e. , how rapidly does the steam generator inventory decrease in relation to the time a high RC pressure trip would occur.

RC system pressure and pressurizer level as functions of tLae are shown in Figures A-2 and A-3, respectively. Reactor trip occurs on high pressure (2300 psig) in 15.8 seconds af ter the loss of one main feedwater pump. Figure A-4 shows the steam generator mass as a function of time and only 307 of the mass is boiled off by the time the reactor trip occurs. This is insufficient inventory decrease to cause a level trip in an anticipatory mode. Figure A-2 shows that no PORV actuation results from this transient.

1494 103

86-1102525-00 Figure A-1 FEE 0 WATER COAST 00tN TO 50%

l 1 1 100 -

90 -

80 - -

70 --

e

~

u E

g 60 - -

B 50 -

40 -

30 - -

20 -- -

10 - -

0 I t I I O 20 40 60 80 100 Time, seconds 1494 104

86-1102525-00 Figure A-2

.FEEDWATER C0ASTOOWN TO 50%

2400; , , , i 2300 -

E

.T a

E a

U oc 2200 -

2100 l

1 2040 I I ' '

O 5 10 15 20 25 Time, seconds 1494 105 7R

86-1102525-00 Figure A-3 FEE 0 WATER COAST 00NN TO 50%

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86-1102525-00 Figure A-4 LOSS OF ONE FEEOPUNP-RAMP TO 50% IN 10 SEC TOTAL S/G MASS, CADDS I I i 1.0 -

O E

.8 -

5 E

= .6 -

.4 i i i i 0 5 10 15 20 25 Tir seconds 1494 107 en

QUESTION

31. Provide full details of the GE model CR-2940 switch to be used to isolate the EFW manual control station from the ICS.

RESPONSE

Full details of the GE Model CR-2940 switch to be used to isolate the EFW manual control station from the ICS was submitted to the NRC on December 3, 1979 (reference letter from GAI providing details from General Electric Co.)

1494 108 Am. 7

QUESTION

52. Provide detailed design features of Fuel Handling Building environmental barrier.

RESPONSE

The following describes the TMI 1 and 2 Fuel Handling and Auxiliary Building

.upply and exhaust systems. The potential leakage paths between butidings or systems and the modifications designed to isolate the unit I refueling floor from the unit 1 Auxiliary Building and from the Control Access Building are discussed. These modifications include ventilation system changes and certain building layout changes. The major ventilation considerations are as follows:

A. The supply and exhaust systems for unit 1 are separate from those of unit 2. However, the unit 1 refueling floor air communicates directly with the unit 2 refueling floor air.

B. The supply systems of the Auxiliary and Fuel Handling Building (FHB) of TMI-1 are separate from each other. Both systems supply air to the building areas through duct distribution systems using outside air drawn from the air intake tunnel. Both supply fans are located in a common tunnel in close proximity to each other.

None of the supply ducts of the Auxiliary Building are located in the FHB area. Thus, there is no potential for air leakage between Auxili-ary and FH Building through outlets or through leaks in the Auxiliary Building supply duct system.

The supply duct main for the FHB serves the general area at elevation 305'-0", the Spent Fuel Cooling Pump area at elevation 305'-0" and then serves the refueling floor at elevation 348'-0". The FHB refueling floor could communicate with the Auxiliary Building through the supply duct system because the general area and the Spent Fuel Cooling Pump area a.e open to the Auxiliary Building through an open doorway at elevation 305'-0".

C. The exhaust systems for Auxiliary and FH buildings of TMI-1 are separate in the specific buildings they serve but the FHB exhaust main becomes common with the auxiliary building exhaust main af ter leaving the FHB.

The common main is directed to multiple filter plenums and fans that exhaust the mixed air.

The building modifications designed to isolate the TMI-1 refueling floor from the TMI-1 Auxiliary Building and from the Control Access Building include (See Drawing 010-006 attached):

A. Two pairs of double doors at elevation 281'-0".

B. An enclosed passage at elevation 305'-0" with twc main doors and one pair of equipment doors.

I494 109 Am. 7

PAGE 2 0F RESPONSE TO QUESTION 52 C. A wall at the east end of the truck bay at elevation 305'-0".

The wall should be removable for large equipment access to the machine shop.

D. A escurity fence at the west end of the dock adjacent to the new enclosure at elevation 305'-0".

E. The stair tower between elevations 299'-2k" & 211'-0" will be modified.

F. Pressure tight doors for the new fuel storage room at elevation 329'-0".

G. Pressure tight doors for the stair tower at elevation 331'-0".

H. An enclosure at elevator entrance with a pressure sealed door.

I. An enc *osure for the ventilation duct chase in the northwest corner of the refueling floor with one p .ir of pressure tight doors.

The TMI-I ventilation system modifications designed to prevent the leakage paths are given below:

A. Air leakage from the FHB through the supply duct, to the de-energized FHB supply fan to the Auxiliary Building are stopped by adding a leak tight damper in the discharge of the FHB supply fan.

B. Air leakage from the FH3 through the supply duct, and the 48" x 24" branch duct, to the FHB general area at elevation 305'-0", and then to the Auxiliary Building will be stopped by blanking off this duct and providing an equivalent opening in the FHB supply duct. This would discharge the required 8000 cfm on the south side of the elevator shaft and this air would rise through the open stairwell to be exhausted at the refueling ficor.

C. Air leakage following the same path as item B above but through a 12 x 12 branch duct in the spent fuel pool cooler area at elevation 305'-0" will be stopped by blanking cff this duct.. The 1000 ct'm exhaust required by this area will then be supplied from the Auxili-ary Building through the wall opening at elevation 305'-0".

D. Air leakage from the Auxiliary Building to the fuel building will be stopped by adding a leak tight damper in the exhaust duct main as it leaves the FHB but upstream of the connection with the Auxiliary Building main (60 x 50, elev. 348).

E. The leak tight dampers added to the FHB supply and exhaust ducts and the supply fan vill function es follows:

1494 110 Am. 7

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1494 111

QUESTION

55. Bulletin 05B Item 1 Your procedure 1102-16, Natural Circulation, includes anticipatory filling of the OTSG prior to securing the reactor coolant pumps.

Submit the analysis performed to provide guidance as to the expected system response.

RESPONSE

The subject procedure requirements are for a smocth and orderly transition to natural circulation. The rate of fill is controlled to maintain plant parameters within normal limits. B&W has provided guidance to the TMI plant staff as to the expected system response. (

Reference:

B&W letter TMI 79-79 dated May 31, 1979, Section 5, is attached).

1494 112 Am. 7

h h Babcock &Wilcox g ,o,n ,,uon o,,u, P.O. Box 1260, Lynchburg Va. 24505 Telephone: (804) 384-5111 May 31, 1979 TMI 79-79 Mr. G. P. Miller Station Superintendent Metropolitan Edison Company P. O. Box 480 Middletown, PA 17057

Subject:

Natural Circulation Operating Guidelines

Dear Mr. Miller:

Actached for your infomation and use is, a copy of Revision 8 of the Operating Guidelines for a Controlled Transition to Natu.al Circulation for Lower Loop Plants. This revision was prepared based on comments received from some of our customers and based on further B&W analysis.

It should be noted that forced circulation is still preferred when available and that if natural circulation cannot be establisF'd and reactor coolant pumps cannot be restored, plant decay heat must '

removed by HPI cooling.

If you have any questions, please advise.

Very truly yours,

p. J. ,=P%

G. T. Fairburn Service Manager GTF:NF , _

Attachment ,

cc: (w/ attachment)

R. M. Klingaman 1404 3 /

11 8 '

J. F. Hilbish L. L. Lawyer J. F. Fritzea R. F. Wilson (3) - GPUSC l R. W. Heward - GPUSC i S. L. Seelinger L. C. Rogers .

S. L. Smith .

The Babcock & Wifecx Company / Established 1867

s * '

n

. OPERATING CUIDE * .

Controlled Transition to Natural Circulation for Lower Loop Plants NOTE: This Procedure is to be utilized in the event there is a need to stop reactor coolant pumps. The use of reactor coolant pumps is preferable to natural circulation for maintaining RCS flow. g

. i .

1.0 Initial Conditions 1.1 Reactor has been tripped for at least 10 minutes i 1.2 One or more RCP's operating l .

i 1.3 Feedwater available .

1.4 HPI available l 1.5 Reactor coolant pressure stable .I 1.6 T is e stable (cooldown rate not to exceed 10*F/hr, no heat up) with steam pressure being maintained by the turbine bypass system .

\

~

2.6 Limits and Precautions

• 2.1 Maintain RCS subcooled -

2.~1'.1 Before RCP's tripped (refer to Figure 1). '

2.1.2 After RCP's tripped (refer to Figure 2).

2.2 Normal cooldown limits should be maintained (including Appendix G EDT limits). . .

2.3 Maintain pressurizer level above pressurizer heaters.

. 2.4 The number of available pressurizer heaters is not limiting for the initiation of natural circulation. However, an effort should be made to maximize the number of available pressurizer heaters to provide the operator with additional pressure control capability. ,

2.5 Prior to stopping the last operating RCP, OTSG 1evel must be established at or above 50% on the operating range and maintained there while on natural circulation.

2.6 The conditions of pressure and temperat:tre allowed by Figure 2 must .

be maintained at all times or RCP operatica and/or HPI cooling' ,

established. -

1494 114 '

3.0 Immediate Actions: None 9

'. C 7. . _ .a- - - - - - --

4.0 Long Term Actions 4.1 Establish subcooling in RCS per Figure 1, using the following as required:

4.1.1 Use available pressurizer heaters. ,

4.1.2 Secondary pressure using turbine bypass system to control Tc.

4.1.3 Use additional makeup.

4.1.4 Initiate HPI if required. -

Caution: If adequate subcooling per Figure 1 cannot be established,

"" continue to run one reactor coolant pump in each loop.

4.2 Establish OTSG level at 50% of the operating range.

Note: Emergency feedwater through the emergency feedwater nozzles

~

is the preferred configuration prior to stopping RCP's.

However, main feed through the main or emergency feedwater nozzles (where available) is permitted provided the OTSG 1evel is established and maintained at or above 50% on the operating range before stopping RCP's.

. 4.3 Establish and control pressurizer level between 100 and 200 inches.

4.4 Stop operating RCP's. ,

Note: The operator must. account for.the loss of pump heat input to the system to prevent overfeeding the OTSG after the RCP's have been stopped.

. ' Caution: Stopping the last operating RCP causes the OTSG 1evel to shift control to the emergency feedwater system and/or the feedwater nozzles. The operator should be prepared to maintain control of feedwater af ter the RCP's are stopped. ,

4.5 Haintain constant or slowly decreasing (cooldown rate noc to exceed 10*/hr, no heatup) Te by controlling steam pressure with the turbine .

bypass system.

4.6, conti uously monir.cr Th. in both loops to assure the pressure and temperature allowed by Figure 2 are maintained. If the conditions of Figure 2 cannot be maintained, go innediately to step 4.9.

4.7 Verify natural circulation by one or more of the following methods.

Note: Indication of natural circulation may not stabilize for 15 to 30 minutes 1494 115

. . .... . . . _ . . . . . . . . _ _ . - 1 .

t

. l

. . - . . . . . . . . . . . . . . . . . . . - . - - - ~ -- - - --

4.7.1 RCS AT increases and stabilizes.

I 4.7.2 Verify heat removal from OTSG's.

a. Turbine bypass valve positions. *

b. Atmospheric desp valve positions. *

c. Feedwater valve positions.

d. Feedwater flow.

Note: May not indicate for low decay heat case.

4.7.3 Incore thermocouple temperatures stabilize.

• 4.8 If natural circulation is confirmed by step 4.7, continue to remove decay heat with natural circulation. -

4.9 If natural circulation cannot be confirmed by step 4.7, maintain the pressure and te=perature limits of Figure 2 or restart ore RCP in each loop.

4.10 If the limits of Figure 2 are exceeded and at least one RCP cannot be started, initiate HPI cooling per the guideline for small break LOCA.

4.11 If feedwater flow is lost: .

4.11.1 Start an RCP.

4.11.2 I= mediately attempt to restore feedrater flow. If feedvater flow is restored, transit' ion to natural circulation.may be y attempted again starting at step 4.2. -

4.11.3, If the limits of Figure 2 are exceeded or OTSG 1evel drops below the low level limit, initiate HPI cooling per the guidelines for small brea'k LOCA. *

~

. . . .-- . . . . . . - . - - -\ -

5.0 Expected Plant Response -

5.1 General -

' This procedure 'is to be utilized in the event that there is a .

need to secure reactor coolant pumps with initial conditions -

as defined in Section 1.0. The principles which form the basis '

of this procedure are: '

1. It is preferred to use reactor coolant pumps to supply core, flow. - - -

2. The hot icg of the RCS must be maintained subcooled to pre-vent formation of a steam bubble which could inhibit natural .

circulation.

1494 116

• mm-

3. With natural circulation or reactor coolant pumps unavailable, the alternative is the EPI cooling mode. 'i .

5.2 Plant response to long term actions 5.2.1 Establishing desired conditions in the RCS Paragraphs 4.1

- through 4.3 will establish conditions in the RCS favorable '

to the initiation of natural circulation. I l,.

- First adequate subcooling must be established in othe RCS

, 7, (per Figure 1) preferably by-raising RC pressure using

. ... pressurizer heaters or by lowering primary te=perature .

. .. using turbine bypass system valves.

l

-' Next, steam generator level must be raised to a level I

known to promote natural circulation. Caution must be U --'

exercised during the feeding of the generators since

"' M IU overcooling may result in a fairly rapid decrease in ~

I '-+ - pressurizer level and RC pressure. The attached Figures h I 3 and 4 provide guidance on the expected system response

' 12 during the feed operation. -

. . J ; . .-

It sheuld be noted that $t may be desirable to raise OTSG Idvel above the 50% level in the operating range. 'Since it

. is known that the 50% level is adequate for initiation of

- natural circulation, this procedure was based on that level.

If the decision is made to raise level above the 50% level caution must be exercised to prevent overcooling the RCS causing loss of RC pressure control or pressurizer. level.

. 5_O The final action prior to tripping the RC pumps is to stabilize pressurizer level. After the pumps are tripped, in2r ..

Th will. increase with Tere=aining essentially constant,

. resulting in a pressurizer level increase.

Y.2Y2 Securing the RC pumps ei t '.

With steam generator pressure being automatically controlled, J.  ! .

Te can be expected to remain essentially constant after*the * .

RC pu=p; tre secured. With the decrease in RC flow, core Th will increase. As a point of reference, a typical in-s crease of 25 - 35*F in T h has been noted for past natural

.' circulation trans'ients. Figure 1 is intended to provide

+ 1494 117 m - - ,- - . . - _ .

^

5 -- T-- -

. , - . ~ . . - , - - ,., .-

adequate subcooling before the RC pumps are secured, so that the hot legs will be at least 20*F subcooled after the pumps are off.

Stopping the last RCP causes the OTSG 1evel control to shift to the emergency feedwater system and/or the emergency feedvater

. nozzles.

5.2.3 Verifying natural circulation Natural circulatica may be verified by: ,

1. Monitoring that Te stays essentially constant with Th increasing and then leveling out. The hot. leg should remain subcooled. ~ ~ ~

2. Monitoring the heat removal from the OTSG's signifying that primary flow is available to remove core heat

- production.

3. Monitoring incere ther=ocouples (if available). The temperatures should rise and then level off and stabi-lize. The final temperature will be a function of many variables including decay heat level. .

~

5.2.4 Contingencies The actions in 4.8 are applicable if feedwater is available.

With feedwater unavailable, the actions in 4.11 are applicable.

In Section 4.11, starting an RCP will provide additional time for the plant operator to try to reestablish feedwater flow by increasing core flow. If feedwater cannot be reestablished.

HPI cooling must be initiated. ,

9 .

1.494 118 .

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' " Figure 2 Allowable Reactor Coolant Outlet f l[

[ and Temperature for Natural j 'l Circulation (All RC Pumps Secured)

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- 1 m x

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                      ._ 18
                   . _ _                                                                                                                                                                                                                   x c           s                                                                                                                                                                                                                          x
      +

w (u = e

                                                                                                                                                                                                                                                   ~

x , t m o ~ _ 6

n.

u ._. ,

e -

w 17 - 150km83.'2fsi/hr

< L w 1- t = - :_- -- [ g __ _ _ _ . s e

• , r O -- ~+

   ,  p              _ _ . _ _ _ .

?! t _

• t

                      -       16 -

=e

                      ~

$5 [ _ .4

• c

-  :. .p. g , _ , _ _ _ _ s g L_______.__ w- E_ _-- ..~-- . . _ _ i r-- = Figure 1. Pressurizer Pressure Vs s

_-______--__ -- Time (f or 144" Pressurizer Level) '

     . ~ -~ _T E ._ . . . _ __ .                       ---. (during Natural Circulation with L=_- ~ 71= -- ~__ Various Heat Losses)                                                                                                                                             s
                               -:1                     =                                        . _ _ _ _ _ _

x

                                                       .__ -- _ _ _ : _ _ ._.- _ _ _                                                                                                          N c . _..              14 . : _. . . _ _ . . _ . . - - - - = . _ . . _ _ _ . _ . ___
                                                               . .__ <._ .. _ _._ _ _.. . -                                                                                                          s

_. _.. . _a

                                                                                                    -~                                                           ~ ~ ~ ~

_0 ---fi ~~~2 = fi- 5 f 4 ; R _.__ L 6 - 8 10 12 - '

                                                                                          -.-=,.____.=;=____--.-----_-----
                                                     '-~
                                                                                    - - :. = =.                                                                                                                   .'                                            ._.

Z _

                                                                                          --.-_---                                             Time, Hours
                                                                                                                             . _. . . _ _ =           :   j-- _._ _ z ______ _                      __

_g___..____ _ - - om o 3- [ D D D y ee - J M.i ei 1494 123

t*'- p . 2 2 _, x L I w

                                                                                               ~-_ ,
                                                                                                                          ~_

21

                                                                                                                                                                  ~ ~

x L ' ~

                                                                                                ---.                                                                                                   ~ ,_

25kw = 14 psi /br= 20 _ ~ h ' E , o E. rs [- 1 3

                                                                                                                                                                                                          ~

E 50kw=26.8 psi /E-~ e b -- v - r 19 b - e ec. 1' ~x N N 's , N s' ~_ i o x x s t ._ es - gg _ -_. . s u._._- ,~

etc x

_ m > A. _ uv u

                                                                                                                                                           ~

_ b g I C x N - - V L- + u E 17 y (C -+- , , m ___. ,s\

. * .W N ia -d

. a. _ % c _. . ?! [L  ; 150kw=68.2 psi /tr r p _, 16 "_,-.

                                                                  ' - - ~

E5 t '~ 'ch b. ---

                                    ~7                                                                                                                                                          .                                                           .

3 E ______ . _ _ _t

                                                                                                                                                                                                  .x EE   p . .__r.--- --                           - - - -
                                                                                     ,               ,                                                                                               N pg    . _ _ _ _ _ _             15 _, __ .                                                                                                                                                               '

4 E - _ _ _ __ _---f--. Figure 2 Pressurizer Pressure _ s

                                              -- Vs Time (f or 2 44" Pren suriz er

[ 3. . ^_ ~ T- .-- '2] __ Level) (during Natural Circulatio- 'N

    =d . EE-i rr --- f or Various Heat Losses)                                                                                                                                                                     'x
                                            +

_, . -x

x. m -m u----- = l . . ._ r _=r- _ _ =_ --- _ ._. . _ _ _ _ _ . _ _ . _ . _ _ 's 14 n _ ___ _ __.. _ ______._._ _

                          -                                               -                       -"                                                                                                                                                  ~

n

                                       ..;.=_=.:.=_=___:.=___=___=_ _ . . _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _

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m x

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g S-E s 9 d- 20 144"=18.5 psi /hr

        ~-

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                    " a       -~

< c - 1 E 2 dm_._._ __

                                                                                                                                                                                                                                               ~

[I g Figure 3 Pressurizer Pressure Vs Time .- j T  :- - -.- , __. with 25 kw Heat Loss for 2

e

.  :. -

Pressurizer Levels (during _ 19 __. _ te  :

• Natural Circulation 7 t-

-3 c, E, .

                             ~

+ +

as _ -=-- = [._.__..__, . . . g y r.._..__,___

                                                          -                                        +-

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J. _ . _.__ . .. . .: _ : : : 2_. ,. ___. =n : ~---_:__.__.22----

                                                                                               . _ . . . _ . _            _                                 .                                                                                  _1_.
                                              ..                      ....__=:.=__                                                                                                                                                                -

f Mr:f f~ . Time, Hours _ .. m; r: _. ._. c i_mg : . _. __ - - g = .. r__. _ _ _ . ._ _3 l j

QUESTION

91. The following additional informatinn is required concerning NUREG-0578 item 2.1.5.B - Hydrogen Recombi ners:

A. You indicate that you have performed hydrogen generation rate calculation in accordance with Regulatory Guide 1.7 in order to verify that adequate time is available following an accident to install the second recombiner and still maintain the containment atmosphere hydrogen concentration with acceptable limits. Submit these calculations for our review. Indicate where the second hydrogen recombiner is to be stored prior to being used and verify that adequate prucc)ures are available and training performed to assure that the second recombiner can be properly and expeditiously installed. B. You indicate that you will perform an evaluation to demonstrate that potential leakage and discharge to t;ie atmosphere of the Intermediate Building air used for recombiner cooling will not result in off-site dose releases in excess of 10 CFR 100 limits. Submit this evaluation for our review.

RESPONSE

A. Je have stated that the design basis for the system is a LOCA with hydrogen generation rates calculated in accordance with Reg. Guide No. 1.7. We have requested one of our consultants, Pickard, Lowe and Garrick, Inc. , to evaluate and confirm that the recombiner which had been purchased for TMI Unit No. 2 is of sufficient capacity for TMI Unit No. 1. Note that the TMI-2 recombiner sizing is based on calculations in accordance with Reg. Guide No. 1.7 which indicate that the 3% by volume hydrogen concentration occurs approximately 250 hours after a LOCA. Since TMI-l and 2 have approxi-mately the same available free containment space and they both have the same size reactors, the time at which containment hydrogen concentrations reach the 3% level are approximately the same. Based on the 250 hours at which time the recombiners are required to operate and the Standard Review Plan Section 6.2.5, Paragraph II.12e requirement, the stored recombiner must be made available to perform its function in a time period that is equal to or less than one-half of the 250 hours. Since the redundant recombiner will be stored in a seismic class I structure on the TMI site and all the connections for its use will be permanently installed, there seems to be no reason why the redundant recombiner cannot be made available to perform its function within 125 hours after a LOCA. A copy of the calculations performed by our consultant is attached. Adequate procedures for the maintenance and installation of the redundant recombiner will be made available to you as soon as they are completed. Training will be performed to assure that the second recombiner can be properly and expeditiously installed. 1494 126 Am. 7

ATTACHMENT TO QUESTION 91 OF SUPPLEMENT 1, PART 2 The attached analysis indicates that any recombiner capable of removing 1 scfm of hydrogen (i.e., about 33 scfm containment air at 3.19. by volume hydrogen) is adequate. The TMI-1 proposed recombiner has a capacity of 57 scfm of containment air and is there-fore adequate. 1494 127 Am. 7

PICKARD. LOWE AND GARRICK. lNC. W., SUITE Gl2  % I 12 o o I ST" sTR E ET, N. \ WASHINGTON, D. C. 2oo36 WAS H I N GTO N, D. C.

    .eA=Cs m.piCaamo wikkaam w. LowC                                                                 TELEPHON U 202 296-8633 B.JOMN GamajCa IRVI N E. CALIFORNI A Twomas m. mosetN S        DOWGLAS C. iDE N JOMN M.vALLANCE           mtCaanD v. CAL Ae nCS C K CITM wCOD A m 3         Smanic AmwCD TMOMA5 C.POTTCn           D CNNIS C. SLEY M Asam Agnau s MICNACL M. SCwwAmT2       A S SOCI ATCS DANIC L w. SteLLwCLL. Jm. STAN MAPLAN ManOLO F. PCm LA November 26- 1979 CAmmOLL L. CATg Mr. D. Slear GPU Service Corporation 100 Interpace Parkway Parsippany, NJ 07054

Dear Mr. Slear:

As requested, we have made calculations required by

-        current NRC licensing practice to establish the acceptable

( sizing of hydrogen recombiners for TMI Unit 1. Our cal-culations are included as Attachment No. 1 hereto. As you are well aware, the NRC is re-evaluating its criteria in this regard, however, should the current requirements hold, these calculations show that any recombiner capable of removing hydrogen at a rate greater than 1 scfm would be adequa1;e. Please call if we can provide further information or clarification. Very truly yours, h/ ' Keith Woodard KW:sm Enclosure cc: Gary Capodanno 1494 128 -

L . Attachment No. 1 (

• Hydrogen Recombiner Sizing for TMI Unit 1 INTRODUCTION The U.S. Nuclear Regulatory Commission currently requires the use of hydrogen recombiners to remove hydrogen as a result of radiolysis following an accident. The procedures and bases for establishing recombiner flow are outlined in Branch Technical Position CSB-6-2 and Standard Review Plan Se.ction 6.2.5.

Following are calculations to support recombiner sizing. HYDROGEN PRODUCED BY METAL WATER REACTION Hydrogen is produced by the reaction of the zircaloy-4 fuel cladding with steam according to: Zr + 2H 2O + ZrO2 + 2H2 1 lb Zr + 0.021978 lb-mole H + 8.4866 scf H 2 2 The total amount of hydrogen produced is based on the amount of reacted zirconium, as determined by the assumptions given in Branch Technical Position CSB-6-2, that is, the amount of H Produced should be five times the maximum amount 2 calculated by assuming that all zirconium in the outside surfaces of the cladding cylinders surrounding the fuel (excluding the clad ing surrounding the plenum volume) has reacted to a depth of 0.00023 inch or that 1% of the total zirconium mass has reacted,'whichever is greater. The mass and surface area of zirconium within the system was taken from the TMI-l FSAR, Volume 1, Chapter 3. Both the surface reaction method (reaction to a depth of 0.00023 inch) 1 1494 129

                                                                        ,,0

and the 1% of total mass method gave approximately the same ( result.' According to the surface reaction method, about 450.4 lbm of Zr is reacted. According to the 1% method, about 471.2 lbm of Zr is reacted. Assuming conservatively that the higher of these values is converted to scf H2 by the above equation and multiplied by five about 20,000 scf of H2 is produced. HYRDOGEN PRODUCED BY CORROSION OF ALUMINUM BY SOLUTIONS (NaOH) USED FOR EMERGENCY COOLING OR CONTAINMEMT SPRAY According to Standard Revieu Plan NUREG-7a/087, Section 6.2.5, the production of hydrogen by aluminhm corrosion should be calculated from a knowledge of the aluminum surface area within the containment and the surface corrosion rate according to the reaction 2Al + 3H 2O + 3H2 + A1 023 1 lb A1 + 0.0555 lb mole H 2 + 19.93 scf H 2 Since the aluminum surface area was not known, it was con-servatively assumed that all aluminum in the system reacted according to the above equation. According to the TMI-1 FSAR, Volume 4, 500 lbm of aluminum is contained within the containment. Therefore, approximately 10,000 scf of H2 gas is produced. HYDROGEN PRODUCED BY RADIOLYTIC DECOMPOSITION OF THE POST ACCIDENT EMERGENCY COOLING SOLUTIONS According to Standard Review Plan NUREG-75/087, Section 6.2.5, the rate of hydrogen production by radiolysis is given by p GE c (t) +GEs s(t) H( (B) (N) 100 t i494 130 2 e

( where

                =      hydrogen production rate, lb-mole /sec SH(t)

P = operating reactor power level, MWt B = conversion factor, 454 gm-mole /lb-mole

                =      Avogadro's number, 6.023 x 10      molecules /gm-mole N

Gc - radiolytic hydrogen yield in core, molecules /100 ev = 0.05 molecules /100 ev E c(t) = gamma ray fission product ene'rgy absorbed by core coolant, ev/sec-MWt Gs = radi lytic hydrogen yield in solution, molecules /100 ev = 0.05 molecules /100 ev E s(t) = energy absorbed in coolant outside core due to fission products dissolved in coolant, ev/sec-MWt The quantity c (t) is defined by: Ev ,(t) = (f Y)c H y (t) where

                   =      fraction of fission product gamma energy absorbed (fY)c           by coolant in core region = 0.10 H (t)     =      gamma energy production rate, ev/sec-MWt Similarly, E (t) is defined by:

E 3(t) = (f g) s H g(t) +f y H7(t) 1494 131 t 3

                                                                         ~*

where

                    =    fraction of total solid fission product (fMS) s           energy absorbed in coolant outside core = 0.01 Hp           =    total solid fission product energy production g(t)         rate, ev/sec-MWt fy           =    fraction of iodine isotope energy absorbed in coolant outside core = 0.5 H y(t)       =    iodine isotope energy production rate, ev/sec-MWt                         -

For calculational purposes, the reactor decay profiles (H (t), H g(t), and H7(t)) specified by the ANS-5.1 draft standard for two-year reactor operation have been fitted by several finite exponential series expressions. The resulting equations are: t H (t) = 1022 (5.1912e-9. x10 t + 0.8743e-6.5x10

                                          ~7                      -8
                       + 0.6557e -5.7x10      t + 0.4098e -7.4x10    t
                                 -8.0x10
                                         -10
                       + 0.150c               t)

Hg = g(t) 2.0 H (t) x t + 0.3279e.1x10 t Hy (t) = 10 2 (0. 8197e-6.1.10

                                           -6
                       + 0,0574e -l.0x10      t) where t      =       time after reactor shutdown (sec) 4 1494 132

( . an be converted to hydrogen production Values for SH (t) rate, VH (t) , in std ft /hr by use of the ideal gas law VH(t)

                        =

1,292,515 SH(t) Values of H (t) , H7 (t) , SH ) ""O H(t) are given for various times in Table 1. The amount of hydrogen accumulated at time, t, is obtained from ., Hydrogen accumulated = / VH( o The integration is carried out numerically for various value's of t in Figures 1 to 4. Figure 5 is a plot of accumulated hydrogen versus time as obtained from Figure 1 to 4. Values of particular importance are Total H Produced Hydrogen Production Rate Time, t 2 (days) by Radiolysis (scf) at Time, t (scf/hr) 12.6 30,000 59.5 20.6 40,000 46.9 SIZING OF RECOMBINER 6 3 The containment free volume at TMI-l is 2x10 ft . The containment will contain 3%'H 2 by volume once 60,000 scf of H2 has been produced and 3.5% H2 jy volume once 70,000 scf of H has been produced. In order to determine when a recombiner 2 would be needed, it was assumed that hydrogen was produced instantaneously by metal-water reaction and by aluminum cor-5 1494 133 O m 3

t ( . rosion and as a function of time according to Figure 5 by radiolytic composition. Results are given in the following table. .. Hydrogen Produced (sef) Hydrogen P: oduction Time Metal-Water Aluminum Radiolytic Rate by Rawiolysis (days) Reaction Corrosion Decomposition Total at Time, t (scf/hr) 12.6 20,000 10,000 30,000 60,000 59.5 20.6 20,000 10,000 40,000 70,000 46.9 Therefore, if the recombiner is to be started when the con-tainment contains 3% H (at atm spheric pressure), it must 2 process about 1 scfm of H 2. If the recombiner is to be started when the containment contains 3.5% H2, it must process about O.8 scfm of H 2. The total flow rates would have to be _ proportionally. larger and account for,, efficiency. . _. ... . 9 1494 134 e 4 ( - 6

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S11 PPT.ESIENT 1, l'AllT 3 1494 146 A:n . 7

1. Your response to questions 36 and 37 does not provide the staff .ith sufficient information to ;ake an evaluation of the high pressure injection (!!PI) design and associated flow rates. '/ie require that you provide the following information:

a. Table of expected IIPI ficw (1 and 2 llPI pumps) in each of the four legs ' ersus RCS pressure (2500 to atmospheric) considering the new eavitating venturi installation.

Provide your analytical / empirical basis for these flow rates. What reduction in flow rate was caused by the inclusion of these flow-limiting devices? Compare these flow rates :a tae n?! tio;. . .t : a a o .n.c - . . . m 'L analysis, " Evaluation of Transient 11ehavior and Small Reactor Coclant System Breaks in the 177 Fuel Assembly Plants," May 7, 1979, Volume 1, Figure 6.2.39.

b. A complete test description for confirmation of adequate flow splits and flows including:

(1) description of temporary flow indications and where they will be installed. (Address why using installed instrumentation is not adequate.) (2) basis for 550 gpm " upper limit" acceptance criteria. (3) range of pressures over which data will be taken. (4) range of installed flow instrumentation. (5) acceptance criteria for flow rates at higher pressures. h'e require that this test confirm that the TMI IIPI design provides adequate flow as assumed in Figure 6.2.59 of the Br.W analysis (above) . Provide your commitment to conduct a test and submit the test procedure which will accomplish this purpose. Response to be provided by January 4, 1979 or sooner, as availabic.

2. Your response to question 36a does not provide sufficient analytical justification for adequacy of the 64/36 flow split for an !!PI line break or your statement that RCS pressure will not expend significant time above 1500 psig for a spectrum of IIPI line breaks. Provide such analyses or confirm that a 70/30 flow split would be achieved and that the existing LOCA analyses are appropriate for a spectrum of IIPI line breaks (between the RCS and the check valve nearest the RCS).

Response to be provided by January 4, 1979 or sooner, as available.

3. Your response to question 36c does not provide the staff with sufficient information to make an evaluation of the cavitating venturi design. Provide justification in the form of test data, calculations, etc., that the cavitating venturis can be relied upon to perform their function for an llPI line break (limit flow out break such that sufficient llPI reaches core). It is our position that the brief test description does not adequately cover the conditions which would

                                                                      ^m. 7 1494 147

result from an HP1 line break. Also, provide detailed drawings, data, and specifications for the cavitating venturis. Response to be provided by 'anuary 1, 1979 or sooner, as available.

4. Item 2.1.7.a of the Lessons Learned requirements states, in part, the following:

The automatic initiation signals and circuits (of the Emergency Feedwater System) shall be designed so that a single failure will not result in the loss of system function. Further review of your proposed design for i.Ph' system has brought into question the capability of the EFh' flow control valves to meet the singic failure criterion in the automatic mode. Our concern is based upon the non-single-failure-proof ICS as the sole source of automatic control signals to the two EFh' flow control valves. (No credit can be taken for the manual control stations in your analysis., Provide a detailed discussion of this aspect of your design that is responsive to the above concern. If conformance to the above require-ment cannot be demonstrated, your response should also include the following: (1) a commitment to upgrade the design to meet this require-ment on an expedited basis, (2) a proposed schedule for completion, and (3) a conceptual design of the proposed modification. The subject of this question is being addressed as described in Supplement 1, Part 2, Question 14 as part of the overall upgrading of the EFh' system. This upgracting should be completed by about January 1981 although effort is underway to shorten the schedule. As design details become available, they will be added to the Restart Report. 1494 148 Am. 7}}