ML19260C736

Amend 10 to Restart Rept,Providing Addl & Nonsubstantive Administrative Changes to Emergency Plan

ML19260C736
Person / Time
Site:	Three Mile Island
Issue date:	01/04/1980
From:	METROPOLITAN EDISON CO.
To:
Shared Package
ML19260C732	List: ML19260C731 ML19260C736
References
NUDOCS 8001080535
Download: ML19260C736 (300)

Text

{{#Wiki_filter:THREE MILE ISLAND UNIT NO. 1 TABLE 2.1-1 List of Isolation signal override Capability Isolation Signal Penetration Reactor HiRh 4 psig 30 psig 1600 psig Line No. Trip Radiation Building Building (SFAS) Break Containment Air Sample 108 N/A N/A C N/A C N/A R.B. Sump 353 C IB C N/A N/A N/A RCDT 330,331 C IB C N/A N/A N/A RCS Sample 328 C IB C N/A N/A N/A R.B. Purge 336,423 C NO NO N/A N/A N/A RCS Letdown 309 A IB C N/A

                                                                                       ,                           N/A     N/A sa Demin Water                         307               C          N/A         C,          N/A        N/A     N/A CD A OTSG Sample                           213, 214          C            IB                    N/A        N/A C,                             N/A NSCCW                              346, 347         N/A         N/A        N/A          NO         N/A     NO U ICCW                                  302, 333,        N/A         N/A        N/A          NO         N/A     NO 334
]O              R.B. Air Coolers                   421, 422         N/A         N/A         C           N/A         C      N/A 7             Core Flood TK                       348,348           C          N/A         C           N/A        N/A     N/A O

Co Legend C = Common Signal Override; initiating isolation condition may still exist. C3 I = Individual isolation signal override capability; procedures governing override to be developed. IB = Individual isolation signal bypass capability

p A = Automatic isolation signal override.

   'ggr                   NO = No override or bypass capability; initiating condition must clear to allow reopening of valve.

T N/A = Not applicable. E' Note: For combinations of initiating signals that are allowable, refer to Table 1 of Appendix A. } m V>

Fage 2 ef 2 THREE MILE ISLAND UN17 g 3 TARI.E 2.1-2 (Cont.) LIST OF CONTA!KMENT ISot.ATION VALVES REQUIRINC MODIFICAtt0NS Valve Velve Line Method Normal Feet Actuel Actuetten Slanal Source Penetretton valve Valve Slee, of Velve Accident Poettien Pseltion 8&W Mo. Service System Tag No. Type in. Ac t uat bn Poeltlen tuleting Hadafied ladiention En lat ing Freposed Modified __ Nat** 334 Reactor Building AN AN-Tl4 Rutter- 48 Air Closed Closed Closed Tee 1,30 1.2.10 8,4,5,10 Outlet Furge fly Line AN-VIS Butter- 48 EMD Closed Closed Closed Tee fly 346 Resetor coolant NS MS-VIS Cate 8 EMD Open Closed Open/ Closed Yes 1,10 8.9.10 7.8,9,10 See Note (1) below Pump Motof Coollag Water Supply 347 Reactne Coolant NS NS-V4 Cate 8 EMO Open Closed Open/ Closed Tee 1,10 8.9,50 7,8,9,10 See Note (I) belou Pump Motor NS-V35 Cate 8 EMO Open Closed Open/ Closed Yes M 8,9,10 7,8,9,10, Coollag Water Return 353 seactor Building WDL WDL-v534 Cete 6 Air Closed closed Clesed Tee 1,10 1,2,10 1,4,5,10 864 does not address Sump Drain WDL-V5 35 Cate 6 Air Closed Closed Closed Yes need on radiat ion olanel 421 Reactor Building R8 RB-V2A Cate 8 g closed open Tee 1,10 8,9 1.2.10; Add aut o init iat ion of JMO Noroel Air Energ, a.B. canlina ca 4 6 Coolers Supply R.R. and 1600 pair R.C. pecesure isoletion signale. s Line N 422 Reactor Sullding BB RB-77 Cate 8 Air Open Closed Open fee 1,80 8,9 Igy; Mermal Air Q Coolers Return M Line s 423 seactor Building AN ANevic Butter- 48 EMD Closed closed Closed Tee 1,4,80 1,2,10 1,4,5,50 Inlet Purge fly

      ^               Line                           AN-710        Sutter-    48         Air        Closed      Closed     Closed           Tee                                                       C fly y

348,349 Core Flood T2. CF CF-V2A&B Clebe EMn Closed Closed Closed M M - 1.5,80 Saeople and N2 Fill -Vl9A&E Cat e E - Lines -720A&B Cat e 1 Air

• Velve Actuation Signal Source

1) 4 reig reactor ialldles pressure teolation 7) Claselfy line to Selsele Category 1

2) 1600 peig (SFAS) leolation 8) 30 pelg reactor building pressure f eelet ton O

3) Redletion aler=, operator action required 9) Line breat teolation signal or protect from ripe uhlp ard jet lapinsement Q

4) Nigh radiat ton (ann-eafety) lootet ton 10) h oote maeval control

5) Reactor trip leolat ion

6) Override capability on individual eelves CQ Notest (1) See emplenet ton in test of TDR - No. TMt-157 pg,10, para IV 1) a) II) and till regarding line break lentat ion.

A Ilne break teolation la not required provided the line een withetend, or le protected from, jet Empingement and the only pipe whly that can break It le the R. C. piping.

2 Manufacturing industries in the res t an pr > duce clothing , v.)>d products, shoes, electrical wiring devices, steel products, packed meat and other food. These activities , vichts a 10-mile radius af the site, are confined chtefly to the cities of Harrisburg , New Cumberland, Steelt 3n, and Midtletsvn. A listing of typical industries within 10 miles of the TMI site and their respective employment is summartzed on Table S. Located within a 5 mile radius af the site are approximately 23 industrial firms that employ about 2,400 persons. There are gas and oil transmission lines located at a minimum distance of appr)ximately 2 miles from TMI. The only power station in the immediate upstream vicinity of TMI is the Crawf ord Station (retired) which is awned by Metropolttan Edison Company. Crawford Station is located about 2 5 miles upstream from the site on the east bank of the Susquehanna River. Immedtately downstream from the site is the York Haven hydroelectric proj ect, consisting af the main dam which averages approximately 10 feet in height and extends about 4,970 feet across the main river channel t3 Three Mile Island; a secandary dam which is about 8 feet high and extends 9 50 feet acrass ths east channel of the river; a pool extending apprutmately 3 5 miles upstream from the dam and contatos about 10,000 acre-feet >f volume; and a head race wall that ts about 20 feet in heig ht extending from the west end of the main dam appraximately 3000 feet to the plant. The York Haven Station is aperated on a "run-of-the-river" basts, and Les power output Ls dependent primarily upon the water ava tlable. However, the reservoir is used for peaking operattan during periods of low.rtver flaw. Brunner Island Station, a large steam-electric generating plant awned by the Pennsylvanta Peswer & Light Company is located on the Susquehanna River approximately ane mile downstream f rom the York Haven proj ect. This station uses water from the river on a "run-through" basis f 3r cooling water. Ihree other hydroelectric generating stations are also located downstream from TMI, with each praj ect having a dam and reservoir on the Susque-hanna River. The three stations are Safe Harbor, Haltwood. and Conevingo Hydr) electric proj ects, located appr aximately 2 5, 31, and 47 miles south af Three Mile Island, respectively. There is also a coal fired, , steam <lectric plant at Holtwood, and the Muddy Run j Pumped St) rage Proj ect is associated with Coneving ) stacLon. The Peach Bottom Nuclear Generating Statt)n is l>cated sling the west bank af the Susquehanna River, about 41 mil's. e downstream >f Three Mile Island, just n >rth >f the Maryland-Pennysivania border and ts the >nly nuclear plant within a 50-mile radlus k] . Am. 10 e

decontamination, etc. In addition, Health Physics Procedures provide instructions on performing surveys, analyzing samples, operating health physics / radiation protection equipment, etc. The pertinent information and details provided in these documents have either been incorporated into the TMI Emergency Plan and/or Implementing Procedures or have been appropriately referenced.

3. A comprehensive set of Emergency Procedures that are used to control plant operations during abnormal and accident conditions have been prepared. Since there is a direct relationship between emergency operations and emergency planning, it is of prime importance that Emergency Procedures and Emergency Plan Implementing Procedures be closely coordinated and complimentary.

Because of this, specific Emergency Procedures will, as appropriate, direct the on-shift operations personnel to the applicable Emergency Plan Implementing Procedure (s). Conversely, Emergency Plan Implementing Procedures will ensure that applicable Emergency Procedures are utilized when appropriate.

4. Metropolitan Edison Company has developed an Emergency Public Information Plan for the TMI Nuclear Station. The purpose of this plan is to describe the methods by which Met-Ed will release information to the media and the public in l addition to the internal dissemination of information in the event of emergencies at IMI. This Emergency Public Informa-tion Plan has been included in the development of the TMI Emergency Plan and is attached as Appendix B.

4.2.2.3 Related County and State Plans The development of the State plans and the TMI Emergency Plan were closely coordinated. In addition, specific State requirements for reporting of emergencies, providing informa-tion and data, recommending protective actions, etc. , have been integrated directly into the Emergency Plan Implementing Procedures. In considering the 10 mile EPZ, there are five county plans (i.e. Dauphin, York, Lancaster, Cumberland, and Lebanon) that have been factored into the development of the TMI Emergency Plan. It is important to point out that not only is the TMI Emergency Plan coordinated with the State and County plans, but the State and County plans are coor-dinated as well. The details of the State and County plans are provided in subsection 4.5.3.1 below. 2-7 Am. 10 1704 116

4.3.0

SUMMARY

OF TMI EMERGENCY PLANNING PROGRAM The TMI Emergency Planning Program, as defined by the Metropolitan Edison Company, consists of two separate but totally coordinated documents. The first document, this Emergency Plan, provides the means for performing advance planning and for defining specific requirements and commitments that will be implemented by other documents and procedures (eg. Administrative Procedures, Surveillance Procedures, Emergency Plan Implementing Procedures, etc.). The second document, the Emergency Plan Implementing Document, provides the detailed information and procedures that will be required to implement the TMI Emergency Plan in the event of an emergency at the TMI Nuclear Station. These two documents are briefly described below. 4.3.1 The Emergency Plan The TMI Emergency Plan assures that all emergency situations, including those which involve radiation or radioactive material are handled logically and efficiently. It covers the entire spectrum of emergen-cies from minor , localized emergencies to major emergencies involving action by offsite emergency response agencies and organizations. The TMI Emergency Plan includes a scheme for classifying emergencies that meets the current guidance (reference 10.20) provided by the Nuclear Regulatory Commission (NRC) . This classification system is described in detail in Section 4.4.0 below. A summary of each classification, its description, purpose and a list of the actions to be taken by the Licensee and Of fsite Authorities is included in Tables 20, 21, 22 and

23. Furthermore, this Plan incorporates specific response criteria (emergency action levels) which will be used in the assessment of emergency situations. Thus, the TMI Emergency Plan provides the overall advance planning required for the development of methods of implementation which will be included in the Implementing Document.

4.3.1.1 In summary, the TMI Emergency Plan provides:

1. A means for classifying emergency conditions in a manner compatible with a cystem utilized by State and County emergency response agencies and organizations.

2. A means of reclassifying such emergency conditions should the severity increase or decrease.

3. Details of normal and emergency operating organizations.

4 General guidelines as well as specific details as to which State, county, and federal authorities and l agencies and other outside organizations are available for assistance.

5. Information pertaining to the emergency facilities and equipment available both onsite and offsite.

6. Guidance for the preparation of detailed Emergency Plan Implementing Procedures. '

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8. Projected river stage 2, 302 f t. at the River Water Intake Structure. l

9. Any earthquake of a magnitude 2,.Olg as indiccted by the

              " Threshold Seismic Condition" annunciator.

10. Transportation of a contaminated, injured individual from onsite to an offsite hospital.

11. Actual or projected hurricane force winds (> 75 mph sus-tained).

12. Onsite aircraft crash outside the protected area fence and not bzpacting on plant structures.

13. Any near or onsite toxic or flammable gas or liquid release which affects the habitability required for normal plant operations.

14. Valid unanticipated alert alarm on an ef fluent radiation l

monitor gas channel. (Monitors RM-A8, RM-A9 and RM-AS)

15. Any fire in a permanent plant structure which cannot be controlled by the Fire Brigade within 10 minutes of discovery or any fire outside plant stuctures requiring offsite firefighting assistance. -

16. Any valid Reacter Building evacuation alarm.

The intent of the values noted above is to provide absolute values which, if exceeded, will initiate the Unusual Event emergency class. In addition to the requirements for declaration of this emergency class that are imposed by the emergency action levels described above, the Unusual Event class can be declared by an action statement in a specific Ersergency Procedure or Alarm Response Procedure. Steps in these procedures state that an Unusual Event has occurred or is occurring and require that an Unusual Event class of emer-gency be declared. All Emergency Plan related actions (notification, etc.) will be carried out in parallel with the remainder of the Emergency Procedure. Lastly, the Emergency Director shall declare an Unusual Event any time that, in his judgement, the plant status warrants such a declaration. Training shall stress the need to analyze all minor events in light of their potential for further degradation of the level of safety of the plant and not hesitate to declare this particular emergency class. 4.4.1.2 Alert The next level of emergency class designated in this Plan is called an Alert. An Alert is the occurrence of an event or series of events that indicate and allow recognition of an actual or potential substantial degradat ion of the level of safety of the plant. As in the case of Unusual vert 1 0z,118 4-3 Am. JO

the Alert class includes emergency situations that are expected to be minor in nature but where it has been deemed prudent to notify more of the offsite emergency participants and mobilize a larger portion of emergency organization. In addition, because of the nature of the Alert class, (releases of l radioactive material possible) broader assessment actions will be started. All of the actions to be taken for each emergency class are described in detail in Section 4.6.0 of this Plan. Events that will initiate an Alert shall be those with the potential of limited releases of radioactive material to the environment. As before, a* situation will only be classified at the Alert level if none of the emergency action levels for a higher class have been exceeded or are expected to be exceeded in the near term. The emergency action levels that shall require an Alert to be declared include (but are not necessarily limited to) the following:

1. Reactor Coolant System pressur, and temperature reach saturation conditions.

2. Reactor Coolant System hot leg temperature 2, 620 F.

3. Reactor thermal power and reacter power imbalance in excess of the safety limits defined by Figure 2.1-2, " Core Protection Safety Limits", in reference 4.10.13.1. l 4 A measured Reactor Coolant System pressure in excess of 2500 psig.

5. Failure of the Reactor Coolant System power operated relief valve to shut (after lifting to relieve pressure).

6. Reactor Building pressure 2,4.0 psig.

7. Reactor coolant total activity > 130 uC1/ml.

8. Primary to secondary system leakage in excess of I gpm.

9. More than one control rod known to be untrippable. ,

10. Loss of all offsite power coincident with loss of both Diesel Generators.

11. Secondary system activity (I-131 equivalent) > 1.0 uCi/ml.

12. Actual river stage 2, 302 ft. at the River Water Intake Structure.

13. Any earthquake of a magnitude 2,0BE levels as indicated by an alarm on the " Operating Basis Earthquake" annun-Ciator.

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14. Tornado warning.

15. An aircraft crash within the protected area or once any permanent plant structure.

16. A fire in any permanent plant structure which requires offsite firefighting capability.

17. A valid count rate on any gaseous plant effluent monitor that would result in a projected dose rate at the exclusion area boundary of > 10 mR/hr (gamma) using adverse meteorology.

18. A valid count rate on any plant iodine effluent monitor that would result in a projected child thyroid dose at the exclusion area boundary of 2, 50 mR in one hour using adverse meteorolorv.

19. 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 2,50 mR in one hour or a whole body dose rate > 10 mR/hr (gamma) using adverse meteorology and building design leakrate.

20. Unanticipated high alarm on any two area and/or process radiation monitors at the same time.

21. A high alarm on the station liquid effluent monitor (RM-L7).

l Again, the values specified are absolute action levels requiring declaration of the Alert level. This class of emergency can also be declared by arrival at an action statement in a specific Emergency Procedure. Steps in these procedures state that an Alert has occurred. or is occurring and require that an Alert class of emergency be declared. All Emergency Plan related actions (notification, etc.) will be carried out in parallel with the remainder of the Emergency Procedure. As in all cases, the Emergency Director shall declare an Alert any time he judges the status of the plant to warrant it. F.e shall specifically consider escalation from the Unusual Event to the Alert class if, in his judgement, the

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situation is not likely to be resolved rapidly or is likely to deteriorate. 4.4.1.3 Site Emergency The next highest level of emergency class designated is the Site Emergency. The Site Emergency class includes accidents in which actual or likely major failures of plant functions needed for protection of the public have occurred. Althouch 4-5 1704 120 Am. 10

9 in reference 4.10.13.2, and (3) any unplanned event at the TMI Nuclear Station which prompts Met-Ed to issue a press release. For the purpose of relating the State's emergency classifi-cation system to the classification system defined in Section 4.4.1 above, a direct correlation between the BRP's Administrative Event and Met-Ed's Unusual Event classifica-tion shall be made. 4.4.2.2 Emergenev Events As stated in the BRP's Plan, an Emergency Event is "any condition or event which has the potential to discharge significant quantities of radioactive material to the public domain. It also obviously includes actual discharge." For the purpose of relating the State's emergency classifi-cation system to the classification system defined in Section 4.4.1 above, a direct correlation between the ERP's Emergency Event and Met-Ed's Alert , Site Emergency, and General Emergency classifications shall be made. l Since both the State's Emergency Event classification and Met-Ed's Site and General Emergency classification's inicude events which have significant potential for radioactive releases, it is imperative that specific guidance for initiating protective actions be available to the " decision-making" personnel in emergency response organizations and agencies. The State has, for planning purposes, adopted the Environmental Protection Agency's (EPA) protective action guides (PAG's) that are speci-fled in reference 4.10.8. It is important to mention that the projected values for dose and dose commitment given as emergency action levels for even the highest class of emergency (i.e. General Emergency) are considerably lower than the EPA PAG's discussed above. Therefore, the declaration of a General Emergency, although an extremely signi-fican event in its own right, should not be construed to mean that the EPA PAG's have, or even will, be exceeded. 4.4.3 Spectrum of Postulated Accidents This section of the TMI Emergency Plan shows that each of the discrete accidents that have been hypothesized for the plant is encompassed within the preceding emergency characterization classes and provides a summary analysis of their implications for emergency planning. 4.4.3.1 Classification of Hypothetical Accidents All of the events hypothesized in Chapter 14 of the TMI Nuclear Station Final Safety Analysis Report (FSAR) fall into one of the four emergency classes outlined above, with approximately half falling into the Alert, Site, and General Emergency categories. Table 6 lists each of these events 4-9 Am. 10 1704 121

and the emergency class that each would be likely to fall into. A complete discussion of any of tk.ese hypothetical events may be found in Chapter 14 of the FSAR. It must be noted that in completing this table the most conse rvative accidents described in Chapte: 14 have been assumed. Therefore, occurrence of some of these accidents (for example with no failed fuel) may not result in as high a class as noted in Table 6. Also, equipment assumed to work in the Chapter 14 analysis were assumed to successfully operate for the evaluation. Failures of required equipment in any of the accident scenarios may result in higher classes of emergencies. 4.4.3.2 Instrumentation Capability for Detection The plant instrumentation that will be used to promptly detect accidents at the plant is discussed in detail in the TMI Nuclear Station FSAR. Table 7 lists each hypothe-tical accident, and the important instrumentation that would be expected to detect each of these accidents. Only major, installed equipment is listed. 4.4.3.3 Mannover and Timing considerations The manpower response and timing considerations for the four emergency classifications are depicted in Table 8. This table includes (1) the number of personnel onsite continuously, (2) the number of personnel to be called to report onsite, and (3) those that are to be called to report to the Off-site Emergency Support Center. Included in Table 8, along with the above information, is the estimated maximum time for the identified personnel to report to their assigned locations. 4-10 Am. 10 1704 122

4.5.0 ORGANIZATIONAL CONTROL OF EMERGENCIES In preparing this section of the TMI Emergency Plan, a slight devi- l ation from the format used in reference 4.10.4 was deemed necessary in order to present the Met-Ed/GPU organizations (normal and emergency) and the various support organizations in a logical manner. It is important to mention that each of the content-related require:ents of reference 4.10.4 for this Section is addressed in the following subsec-tions. To facilitate the review of this section against reference 4.10.4 criteria the following cross index should be used: o Subsection 4.5.1.1 below was prepared in accordance with subsection 5.1 of reference 4.10.4. o Subsection 4.5.1.2 below was prepared in accordance with subsection 5.2 of reference 4.10.4. o subsect ions 4.5.1.3, 4.5.1.4 and 4.5.1.5 below were prepared in accordance with subsection 5.3.1 of reference 4.10.4 o Subsection 4.5.2 was prepared in accordance with subsection 5.3.2 of referenci 4.10.4 o Subsection 4.5.3 sas prepared in accordance with subsection 5.4 of reference 4.10.4 4.5.1 Licensee Organizations The TMI Generation Group is the organization 'which operates and provides technical support for the TMI Nuclear Station. This organi-zation is staffed by Met-Ed/GPU personnel. Ihe following five subsections provide a detailed description of the TMI Generation Group Station and Station Support Organizations during normal operations and, in addition, a detailed description of the onsite and offsite emergency organizations that can be activated from the normal organizational arrangements. A description of a basic organization for long-term recovery operations is also provided. 4.5.1.1 Normal Station Organization

1. A block diagram of the TMI Nuclear Station Organiza-tion is provided as Figure 9. The diagram illustrates the levels of responsibility within the station organization. The personnel staffing the normal station organization are usually onsite from about 8 AM to 5 PM during the normal work week (i.e., weekends and holidays excluded).

a. The Vice President Met-Ed has the overall respon- l sibility for the day-to-day operation of Unit 1.

In carrying out his responsibilities he directs the management staff in the execution of their functional responsibilities.

b. The Manager Unit 1 is responsible for the overall operation and maintenance of the Unit. He reports directly to the Vice President of Met-Ed and is directly responsible for operating the unit in a safe, reliable and efficient manner.

5-1 Am. 10

                                                                             }Q4 }2)

c. The Manager Plant Engineering is responsible for the coordination of the cechnical engineering staff, chemistry and Shift Technical Engineers.

The Manager Plant Engineering also reports to the Vice President Met-Ed.

d. The Manager Training reports to the Vice President Met-Ed and is responsible for the conduct of all training functions .

e. The Manager Administration and Services is respon-sible for planning, organizing and directing the l day-to-day activities of the Personnel, Budgets and Reports, Security, Facilies, Document Control, Office Management and Safety functions. This position reports to the Vice President Met-Ed.

f. The Radiological Controls Manager is responsible for all matters affecting the Radiation Protection and Health Physics aspects directly associated with-the operation and maintenance of Unit 1. This position reports to the Vice President Met-Ed.

2. The Operations Department provides operators onsite on a rotating shift basis to ensure the safe and proper operation of the plant 24 hours per day, 7 days per week. In addition, personnel from other departments within the station c ganization are also assigned to shifts to provide adcitional capabilities.

Requirements for minimum shift crews are specified in Section 6.2 of the Technical Specifications (reference 4.10.13.1), however, the typical TMI Unit I shift organization is shown on Figure 10. It is important to mention that the shift organization can be augmented, in the event of an emergency, with off-duty personnel within 60 minutes.

a. The Shift Supervisor, who is on duty at all times, is in the immediate onsite position of authority and responsibility for the safe and proper operation of the plant. The Shift Supervisor will be responsible for the initial evaluation of any abnormal or emergency situation and for directing the appropriate response. If it is determined that an emergency exists, those responsibilities assigned to the Emergency Director will be assumed by the Shift Supe rvisor.

Under such circumstances, the Shift Supervisor will initiate appropriate actions, implement proper procedures, notify appropriate offsite emergency response organizations and agencies (i.e. Dauphin County, PEMA, NRC) and the Duty Section Superintendent, and retain such respon-sibilities until relieved as the Emergency Director. The Shift Supervisor shall, during normal and emergenev operations, maintain control over plant operations as the senior 1704 124 5-2 Am. 10

h. Ensure that adequate protective measures are taken by personnel performing emergency efforts. This includes those individuals assigned to the Emergency Control, Technical Support, and Operations Support Centers.

i. Ensure that accurate exposure records are maintained.

j. Review and evaluate updated information and data.

k. Relay significant information and data to onsite and offsite organizations, agencies, and response teams.

1. Determine the necessity for onsite evacuation.

2. Plant Staff Emergency Assignments

a. Communicator The Communicator will report to the Emergency Director. He will function as liaison between the Emergency Director, offsite organizations and agencies, onsite organizations and agencies, and the onsite emergency organization (i.e. Technical Support Center Coordinator, " oup Leader - Technical Support , NRC-Bethesda, and the Babcock and Wilcox Company). As required, the Communicator will provide, using available equipment, relicble and accurate communications in accordance with the appropriate Emergency Plan Implementing Procedures.

In addition, he is responsible for maintaining records of outgoing and incoming communications. He will have designated assistants that will be utilized in an emergency as necessary.

b. Technical Support The Technical Support Center Coordinator and his staff of engineers will report to the Emergency Director. The technical support personnel will analyze current and projected plant status and, through close communications (via the Communicator) with the Emergency Director, provide technical support and recommendations regarding emergency actions. In addition, the Technical Support Center l Coordinator will, as necessary, provide a direct interface with the Group Leader - Technical Support.

More specifically, they will: (1) Analyze mechanical, electrical, and instru-ment and control problems: determine alter-nate solutions, design and coordinate the installation of short-term modifications. (2) Analyze thermohydraulic and thermodynamic problems and develop problem resolutions. (3) Assist in the development of Emergency Procedures, Operating Procedures, etc. as necessary for conducting emergency operations. (4) Analyze conditions and develoo guidance for the Emergency Director and operations personnel e 5-6 Am. 10

(5) Resolve questions concerning Operating License requirements with NRC representatives,

c. Plant Operations (1) The Operations Coordinator is responsible for coordinating operations and maintenance activities through the Shift Supervisor and the Emergency Maintenance Coordinator. The Operations Coordinator may not relieve the Shift Supervisor of or specifically direct plant operations unless he is a licensed Senior Reactor Operator. The Operations Coordinator will report to the Emergency Director.

(2) The TMI Unit 1 Operations Shift, under the direction of the Shift Supervisor, is respon-sible for the safe and proper operation of the plant at all times. Therefore, the operations shift will respond to all abnormal and emergency situations and take action as necessary to improve and/or terminate any accident. - The shift organization will be self-reliant for a sufficient period of time to allow for the notification of required personnel and for them to assemble and integrate into the emergency organization. To ensure the shift can respond and function in an emergency, the Shif t Supervisor is responsible for the . initial assessment and evaluation of the situation and will initiate the necessary immediate actions to limit the consequences of the accident and bring it under control. In addition, he will assume and carry out the responsibilities of the Emergency Director until relieved by the member of the TM1 Staf f assigned this duty. The Shift Technical Engineer, as discussed in subsection 4.5.1.1 2.e. above, will advise the Shi f t Supervisor on activities that impact the safe and proper operation of the plant. The shift organization personnel are familiar with the operation of plant systems and the location and use of emergency equipment. Some members of each shift are trained in fire fighting, first aid, and the use of radiation monitoring equipment. In addition, a Radiation / Chemistry technician is assigned to l each shift to provide. 24 hour per day, 7 day per week coverage. 1704 126 5-7 Am. 10

engineer, consultants, the Nuclear Regulatory Commission, and other individuals shall be included in the Technical Working Group as appropriate.

6. The Task Management / Scheduling Group sets priorities; develops plans and schedules; coordinates and monitors l the status of tasks; and reports the work progress of all the technical groups. In addition, the group provides liaison with the Nuclear Regulatory Commission.

7. The Technical Support Group is responsible for providing engineering support, technical planning and analysis, procedure support, control room technical support, data reduction and management, and support relating to licensing requirements.

8. The Plant Operations Group consists of the plant staff with substantial augmentation from other organizations.

This group is responsible for performing all plant operations and maintenance activities, limiting and controlling personnel exposures, in plant health physics management, terminating or mini:nizing offsite releases, stabilizing plant conditions, and restoring the plant's ability to function nomally and respond to any further emergencies.

9. The Waste Mangement Group is responsible for safely and effectively managing the quantities of radioactive gases, liquids, and solids that might exist during the initial phases of recovery. Subsequently, this group is responsible for the development and impicmentation of short and long-term plans to manage and process contaminated solids, liquids, and gases; quantifying the degree of contamination of buildings and systems:

and the establishment of processing priorities based on plant needs.

10. The Plant Modifications Group is responsible for providing the engineering, design, materials and construction necessary to complete required modifications to plant systems, equipment, and structures.

11. The Industry Advisory Group is designed to function in parallel with the other technical groups and is not intended to be part of the implementation structure.

The purpose of this technical group is to objectively look into potential problems, maintain a current awareness of the perceived plant and reactor core status, and provide independent assessment based on exnerience and judgement rather than detailed engineering review and calculations. 5-15 1704 127 Am. 10

4.5.2 Local Services Sueoort

       'lhe nature of an emergency may require augmenting the onsite emergency organization therefore, it may become necessary to request and utilize assistance furnished by local personnel, organizations, and agencies.

Since it is essential that support from local law enforcement agencies, fire departments, hospitals, physicians, and ambulance services be available on relatively short notice, agreements, copies of which have been included in Appendix C, have been made with the following personnel, organizations, and agencies: 4.5.2.1 Medical Sunoort Organizations and Personnel

1. Hershey Medical Center

2. Physiciaas

3. Radiation Management Corporation 4 Bainbridge Fire Company (ambulance service)

5. Liberty Fire Company (ambulance service)

6. Londonderry Township Fire Company (ambulance service)

7. Rescue Hose Company No. 3 4.5.2.2 Firefighting Organizations

1. Bainbridge Fire Company

2. Liberty Fire Company

3. Londonderry Township Fire Company 4 Rescue Hose Company No. 3 4.5.2.3 Law Enforcement Agencies

1. Pennsylvania State Police

2. Middletown Police Department 4.5.2.4 Miscellaneous Supoort

1. Consolidated Rail Corporation 4.5.3 Coordination with Government Agencies Metropolitan Edison Company has and will continue to work closely with State, county, and federal agencies in coordinating emergency planning activities for the Emergency Planning Zone in order to ensure the health and safety of the general public. As a part of this coordination, each participating agency has been assigned specific responsibilities and authority for both emergency planning and emer-gency response. Also as a pa,rt of this combined effort, speci fic emergency-related notification and information reporting requirements be: ween Met-Ed and the various participating agencies have been defined.

Additional reporting requirements, which are specified in reference 4.10.14 and in Section 6.9 of reference 4.10.13.1, will also be met. A brief description of the key elements of each of the participating State, county, and federal agencies is provided in the foIlowing subsections. Additional information pertaining to emergency related notification requirew nts that activate the emergency response organi-zations and the subsequent information reporting requirements is provided in Section 4.6.1. 5-16 Am. 10

4.5.3.1 State Agencies The Pennsylvania Emergency Management Agency (PEMA) was established under the Commonwealth of Pennsylvania, Public Law 1332, and is in general, responsible for coordinating emergency services in the State. In order to plan for emergencies in a manner that compliments federal guidelines and implements State directives that are related to emergency operations, the Commonwealth of Pennsylvania Disaster Operations Plan was prepared and issued under the authority of. and in accordance with the provisions of Public Law 1332. l Annex E, Emergency Nuclear Incidents (Fixed Nuclear Facili-ties), to the Disaster Operations Plan provides guidance to the State, county, local, aad federal agencies and nuclear power plant facilities in the development and implementation of emergency plans associated with radiological emergencies. Annex E is attached to this Plan as Appendix D.

1. Pennsylvania Emergency Management Agency The Pennsylvania Emergency Management Agency (PEMA) has been assigned as the lead State agency for the coordin-ation of Radiolological Emergency Response Plans prepared by the State, County, and Local governments and the TMI Emergency Plan prepared by Met-Ed. In addition, should a radiological emergency occur at the TMI Nuclear Station that requires the implementation of State, county, and local government Radiclogic al Emergency Response Plans, the State agency through which the Governor will exercise coordination / control will be PEMA. However, as in all emergencies, the Governor retains directional control. The Director of PEMA has the overall responsibility for the agency's operation. To ensure that PEMA can provide 24 hour per day, 7 day per week response to emergencies, PEMA maintains an off-hours Duty Office. Specific respon-sibilities assigned to PEMA are defined in Section IX of Annex E to the State's Disaster Operations Plan which is attached as Appendix D.

To ensure that PEMA can adequately respond to, and cope with a radiological emergency and provide the necessary interfaces and support, the Department of Environmental Resources, Bureau of Radiation Protection will, in an advisory capacity, work with PEMA.

2. Department of Environmental Resources The Department of Environmental Resources (DER), under the administration and technical direction of the Secretary, DER, is generally responsible for gathering and evaluating technical information and for supplying such information and technical advice and recommenda-tions to PEMA and the State Council of Civil Defense.

5-17 Am. 10 1704 129

Specific responsibilities assigned to the DER are defined in Annex A and Annex E to the State's Disaster Operations Plan. The Commonwealth of Pennsylvania will conduct, on a continuing basis, an education program designed to familiarize the public with factual information concerning nuclear radiation. (Appendix D Section VIII E. ) The Department of Education will coordinate with the Department of Environmental Resources in the development of programs on nuclear radiation and assist DER and Health in the development of an enhanced public information program concerning the effects of nuclear radiation. Within the DER, the Bureau of Radiation Protection (BRP) has been delegated DER's responsibilities as they apply to radiological emergencies. The Director, BRP, is responsible for meeting .these delegated responsi-bilities, however, the Secretary, DER, will maintain overall responsibility. Specific responsibilities assigned to the DER /BRP that are appropriate to radio-logical emergencies are defined in Section IX of Annex E to the State's Disaster Operations Plan which is attached as Appendix D. To provide for emergency response capability, the BRP has made provisions for 24 hour, 7 day per week inter-face with PEMA. 4.5.3.2 County Agencies Public Law 1332 states that "Each political subdivision of this Commonwealth is directed and authorized to establish a local emergency management organization in accordance with the plan and program of the Pennsylvania Emergency Management Agency. Each local organization shall have responsibility for emergency management, response and recovery within the terri-torial limits of the political subdivision within which it is organized and, in addition, shall conduct such services outside of its jurisdictional limits as may be required under this part." Therefore, each County and Local Emergency Management Coordinator in the State is responsible for establishing a civil defense organization within their respective jurisdic-tion, developing plans and preparing for emergency operations in conformity with the State's with the State's Disaster Operations Plan and Public Law 1332 . With respect to the 10 mile EPZ, the 5 counties identified below have prepared emergency plans that are coordinated not only with the State's Disaster Operations Plan but with the TMI Emergency Plan as well. Local government plans are either included directly with their respective County's plan or are maintained as separate, but coordinated documents. 5-18 Am. 10

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associated with the incident and verifying that emergency plans have been implemented and proper agencies notified. Should NRC personnel be dispatched to the scene they will, as needed, assist, in coordin-ation with Radiological Assistance Teams provided by the Department of Energy, in providing to State and local agencies advisory assistance associated with investigating and assessing radiological hazards to the public. To ensure reports can always be made, the Region I NRC Office of Inspection and Enforcement (I&E) is equipped with a 24 hour emergency telephone number and has assigned duty officers. The Region I I&E Office is located in King of Prussia, Pennsylvania. A lettter of understanding is provided in Appendix C.

2. Department of Energy The Department of Energy (DOE) has established reveral regional coordinating offices for the purpose of providing radiological emergency assistance to State, county, and local governments; nuclear power plant facilities; and other federal agencies.

To ensure 24 hour, 7 day per week emergency response capability, the DOE has established a continuously manned Security Operations Center in each of their designated regions. The Security Operations Center for the region in which the TMI Nuclear Station is located is in Upton, New York, and the Brookhaven Office. A l letter of understanding is provided in Appendix C.

3. U.S. Coast Guard During a radiation emergency which could involve exposure to offsite personnel, the U.S. Coast Guard will provide assistance by maintaining traffic control on the Susquehanna River. By authority of the Pennsyl-vania Emergency Management Agency, the U.S. Coast Guard, as assisted by the Coast Guard Auxiliary, may also be used to keep unauthorized personnel from entering Three Mile Island and adjacent islands until emergency conditions are controlled. Assistance may be obtained by calling the 24 hour emergency telephone. A letter of understanding is provided in Appendix C.

4 Department of the Army The 56th Ordinance Detachment of the Department of the Army will provide Explosive Ordinance Disposal personnel in response to requests for assistance during bomb threats. Additional details of this arrangement are specified in a letter agreement which is provided in Appendix C. 5-21

                                               ^= tol 7 04  131

5. Federal Aviation Administration TBD - A letter of agreement with the Federal Aviation Administration is being pursued.

6. National Weather Service TBD - A letter of agreement with the National Weather Service is being pursued.

4.5.3.4 Interagency Communications Section 4.6.1 and 4.7.5 describe the communications con- l cepts and equipment that will be utilized by TMI emergency response personnel. The Communications section of the State Disaster Operations Plan (included in Appendix D) and Section XI, "Communica-tions", of the Department of Environmental Resources Bureau of Radiation Protection Plan for Nuclear Power Generating Station Incidents (included in Appendix D) describe the systems and communication concepts that are utilized by the Commonwealth of Pennsylvania in order to coordinate emergency activities. The Operations Officer, and after normal hours the Duty Officer are the primary contacts for PEMA. The Emergency Management Coordinator, his Deputy or Communications Officer are the primary contacts at the county level. The Emergency Management Coordinator at the local level is that political subdivision's principal contact. The Director of the Bureau of Radiation Protection or his designated alternate is the primary contact for the Depart-ment of Environmental Resources. 4.5.4 Training of State, County and Local Governments The Pennsylvania Emergency Management Agency (PEMA), in accordance with Annex E of the Disaster Operations Plan, will conduct and parti-cipate in annual training exercises that involve State, county and local government agencies and consist of (1) scenario development, (2) reviewing Radiation Emergency Response Plans, procedures, and response capabilities, (3) testing of communications, radiological monitoring instrumentation and warning systems, and (4) critique of the exercises. PEMA also provides a comprehensive series of courses consisting of home study, seminars / workshops, and formal training programs on subjects such as basic Civil Defense, basic health physics, operation of radiological survey instrumentation, emergency management, and county coordinator development courses and conferences. These programs are utilized to provide state emergency monitoring teams with the necessary skills to perform their functions, to provide new County Civil Defense Directors / Coordinators with the necessary knowledge to implement radiation emergency response plans and procedures, and to instruct local government of ficials, their staffs, and other key personnel in emergency preparedness planning. 1704 132 5-22 Am. 10 O

4.6.0 EMERGENCY MEASURES This Section identifies the specific measures that will be taken for g each class of emergency defined in Section 4.4.0 of this Plan. The I logic presented in this Section is used as the basis for detailed Emergency Plan Implementing Procedures which define the emergency actions to be taken for each emergency class. Emergency measures all begin with (1) the recognition and declaration of an emergency class, (2) notification of the applicable agencies for each emergency class, and (3) mobilization of the appropriate portions of the emergency organization. The additional measures are organized into assessment actions, corrective actions, protective actions and aid to affected personnel and are described in the sections below for each emergency class. 4.6.1 Activation of Emergency Organizations Meeting or exceeding a predetermined value or condition specified as an emergency action level in.an Emergency Plan Implementing Procedure shall require the implementation of that proc edure . Specific emergency action levels for each emergency class (i.e. Unusual Event, Alert, Site Emergency, and General Emergency) are defined in Section 4.4.1 above. The Shift Supervisor, in implementing the Plan will initially classify the emergency, ensure that required notifications are made, and notify the Duty Section Superintendent or his designated alternate. The Duty Section Superintendent will, working closely with the Shif t Supervisor, perform an overall assessment of the emergency in order to determine its most appropriate classification and, based on this determination, activate larger protions of, or the entire emergency organization. A more detailed discussion of the methodology that is used in activating the emergency organizations during each class of emergency is provided below. In addition Figures 15, 16, and 17 provide a visual display of the communications networks that have been planned for notification requirements, information reporting, and decision making with respect to taking protective action for the public , respectively. 4.6.1.1 Shift Foreman / Control Room Ooerators

1. Should emergency situations (real or potential) arise, it is expected that the Control Room Operator (s) and/or the Shift Foreman will be initially made aware by alarms, instrument' readings, reports, etc. The Control Room Operator (s) shall ensure that the Shift Foreman and/or the Shift Supervisor are immediately informed. The Shift

• Foreman shall, if not already accomplished by the Control Room Operator (s), inform the Shift Supervisor immediately.

4.6.1.2 Shift Supervisor

1. The Shift Supervisor, when informed of an emergency situation, is responsible for performing the assessment of the emergency (e.g. plant systems and reactor core status, radiological conditions, etc.) in the fo llowing manner:

6-1 Am. 10 1704 133

(3) For a Site Emergency, the entire onsite emergency organization shall be activated. A major portion, if not all of the offsite emergency organization shall also be activated. (4) For a General Emergency, the entire onsite and offsite emergency organizations shall be activated,

c. Inform the support member of the Duty Section of the emergency and direct him to activate the offsite and/or onsite emergency organizations as determined immediately above.

d. Report to the Emergency Control Center and assume the position of Emergency Director in the onsite emergency organization.

NOTE The Shift Supervisor shall assume the responsibilities assigned to the Emergency Director until properly relieved by the Duty Section Superintendent. 4.6.1.4 Emergency Director

1. The Emergency Director shall locate himself in the Emergency Centrol Center and assume the responsibili-ties specified in paragraph I of subsection 4.5.1.3 above.

4.6.1.5 Dauphin County

1. The dispatcher at the Dauphin County EOC shall notify the Dauphin County Emergency Management Coordinator I or his designated alternate.

2. The Dauphin County Emergency Management Agency (EMA) {

shall, through area and local civil defense personnel and other means, notify local municipalities with priority favoring those nearest TMI. 4.6.1.6 Pennsylvania Emergency Management Agency

1. The PEMA Duty Officer or Operations Officer at the l State EOC shall, upon receiving notification of an emergency at the TMI Nuclear Station, immediately notify the Bureau of Radiation Protection.

2. PEMA shall, in accordance with standard operating procedures, notify the following personnel, organi-zations, and agencies as appropriate:

a. Dauphin County EMA

b. Other affected county emergency management organiza-tions (e.g. York, Lancaster, Cumberland, Lebanon)

c. Other affected s:ates (i.e. Maryland)  !

6-3 Am. 10 1704 134.

d. Selected State agencies such as the:

(1) Department of Environmental Resources (2 ) Department of Agriculture (3) Department of Health (4) State Police (S) Department of Transportation

e. Selected Federal agencies.

4.6.1.7 Bureau of Radiation Protection

1. The Bureau of Radiation Protection employee (i.e.

Incident Manager) who receives the notification from PEMA that an emergency exists at the TMI Nuclear Station shall:

a. Call the TMI Nuclear Station Emergency Control Center to:

(1) Determine the classification of the emergency (2) Obtain information and data pertaining to the emergency.

b. Initiate activation of the BRP emergency response organization. ,

c. Advise the PEMA Duty Officer or Operations Officer l of the BRP's initial assessment of the emergency.

d. Notify selected Federal agencies.

4.6.2 Assessment Actions Effective coordination and direction of all elements of the emergency organization requires continuing accident assessment throughout an emergency situation. Each emergency class will invoke simil.ar assessment methods, however, each class imposes a different magnitude of assessment effort. In the following subsections assessments actions to be taken for each emergency class are outlined. 4.6.2.1 Assessment Actions for Unusual Events The detectien of an Unusual Event will arise from either exceeding a specific emergency action level for this class (see Subsection 4.4.1.1) or by an action statement in an Emergency Procedure. Detection of the event in either case will come as the result of alarms, instrument readings, . 6-4 Am. 10 1704 135

recognition through experience or a combination thereof. The continuing assessment actions to be performed for this class of emergency will be in accordance with the Emergency Plan Implementing Procedure for an Unusual Event and will consist of the normal monitoring of Control Room and other plant instrumentation and status indications until the situation is resolved. If a fire was the reason for the declaration ot an Unusual Event, the Fire Brigade leader who will have reported to the fire location will make continuing assessments based on his experience and report to the Emergency Director on whether offsite firefighting assistance is required. 4.6.2.2 Assessment Actions for Alerts Once an accident has been classified as an Alert by the Emergency Director, assessment actions will be performed in accordance with the Emergency Plan Implementing Pro-cedure for an Alert. These actions will include:

1. Increased surveillance of in-plant instrumentation.

2. If possible, the dispatching of an Emergency Repair Team to the identified problem area for confirmation and visual assessment of the problem.

3. The dispatch of onsite radiological monitoring team (s) to monitor for possible releases and provide rapid confirmation of correct accident classification.

4 Dose Assessment -- If a radiological accident is occuring, surveillance of the in plant instrumentation necessary to obtain meteorological and radiological data required for calculating or estimating projected doses will commence. The dose assessment activity will continue until termination of the emergency in order that updating of initial assessments may be provided to all concerned offsite agencies and to the Emergency Director. Emergency Plan Implementing Procedures are provided to allow rapid, consistent projection of doses. 4.6.2.3 Assessment Actions for Site Emergencies The assessment actions for the Site Emergency class are similar to the actions for an Alert, however, due to the increased magnitude of the possible release of radioactive material, a significantly larger assessment activity will occur. The necessary personnel for thiz assessment effort will be provided by mobilization of the entire onsite and of fsite emergency organizations. Specifically,

1. An increased amount of plant instrumentation will

             'be monitored. In particular, indications of core status (e.g. incore thermocouple readings) will be     l monitored closely.                                     [

6-5 Am. 10 1704 136

4.6.3 C,orrective Actions Detailed operating procedures are available to the operators for use during emergencies as well as normal operations. Specific Emergency Procedures are provided to assist the operators in placing the plant in a safe condition and taking the necessary supplemental corrective ac t ions . In addition, operations personnel are trained in the operation of plant systems and their associated procedures and will, therefore, be capable of taking appropriate corrective actions. Selected TMI Nuclear Station Staff personnel, including operations, health physics, and maintenance personnel are trained and assigned to emergency teams. These teams will be able to respond as set forth in the Emergency Plan Implementing Procedures in order to assess condi-tions and take any available corrective actions. Maintenance ner-sonnel will provide the necessary crafts expertise to effect repair and damage control functions. Corrective actions will normally be planned events that are taken to ameliorate or terminate the emergency situation. Planned radioactive releases or corrective actions that may result in a radioactive release will be evaluated by the Emergency Director and his staff as far in advance of the event as is possible. Such events and data pertaining to the release will, if at all possible, be reported to the appropriate offsite emergency response organizations and agencies. 4.6.4 Protective Actions Protective actions are emergency measures taken during or after an emergency situation that are intended to minimize or eliminate the hazard to the health and safety of the general public and/or site personnel. Such actions taken onsite are the responsibility of Metropolitan Edison Company while those taken offsite fall under the jurisdiction of the Commonwealth of Pennsylvania and other offsite response agencies. All visitors to the site will be either escorted by an employee knowledgable as to emergency plan response or will receive training on actions required by them during an emergency. 4.6.4.1 Protective Cover, Evacuation. Personnel Accountability During an emergency, the relocation of persons may be required in order to prevent or minimize exposure to radiation and radioactive materials. The following subsections discuss the policies applying to such situ-ations.

1. Plant Site All persons onsite at the time of an Alert, Site or General Emergency who do not have emergency assignments (i.e. non essential personnel) shall be notified of l

the emergency class by announcement over the public address system. These persons will be instructed to report to assembly areas for accountability, mon-itoring and possible evacuation. Such persons will 6-7 Am. 10 1704 137

following are the exposure guidelines f or these emergency activities. Ltfe - saving action - 100 Rem Correcetve action - 2 5 Rem Personnel involved in any of the above actions must be volunteers. The Emergency Director shall authartze the above exposures and is responsible f sr maintaining expo-sures below these values. He shall, if possible, seek advice from the Radiolog teal Assessment Coordinator. He shall assure that all measures are taken to minimize )ther exposures (such as internal exposure) during these activt-ties. 4652 Decontaminattan and First Aid Decantamination materials, specialized equipment and suppites , and portable first aid kits are available in the Decantamination Area. Portable health physics instruments for personnel monttoring and portal monitors are available at the access contr21 point. Decontamination showers and a sink, both sf which drain to the radvaste system, and an eyewash are also located in the Decontamination Area. All personnel leaving the controlled area vill be non-itsred for contamination by use of portal monitors, and/or hand and foot counters, and/or friskers. During emergen-ctes all personnel onsite will, as necessary, be checked for contamination. f Personnel found to be contaminated vill undergo decon-tamination by health phystes personnel (or other qualif ted personnel as specified in Health Phystes Procedures). It is preferable that personnel decontamination be perf ormed by health phys tes personnel, however, other Met-Ed personnel are instructed in both decontamination and first aid pracedures. Measures will be taken to prevent the spread af contamination. Such measures may include isolating affected areas, placing contaminated personnel in " clean" protective clothing before moving, and decontaminating af fected personnel, their clothing, and equipment prt)r to release. Emergency first aid and medical treatment will be g tven to injured personnel who are contaminated. Shtf t personnel, tratned in first ald, will be available .ansite on a 24 hour per day basis and will assist contaminated personnel either at the scene of the accidant .)r in the first aid room located in the 30 Y elevation of the Service Butiding. Pryy tstons have been made, thraugh agreements, to ensure contaminated and injured personnel vill receive specialized medical treatment if necessary. The Hershey Medical Center has agreed to accept c-)ntaminated patients f-ar emergency medteal and surg teal treatment and/ 3r >bservatt an. I 6-12 Am. 10 1704 I38

     )     A pulse geaeratar or current s 2urce is used f or electrically checking each mantt)r 3r sub-system. Electrical input tests measure the functional aperation af the monitoring system fram the detector output thru the readsut devices.

2 The modules are designed so that an alarm and/or indication is intelated when fatlure accurs anywhere in the channel.

a. Area Radiation Monitoring The area radtation monitoring subsystem is comprised if channels which uttitze an tanizattan chamber type detector housed in a weatherpraaf contatner and equipped with a remote controlled check source. The local alarm and readsut for each of these channels is separate from the detector and is also housed in a weatherpr33f container.

One channel is pravided to monitor the high level of radiation that would be characteristic of the post-accident atmosphere in the containment. The detector is densensitized by a lead shield. The Control Room readout modules are located in the radiattan monitoring penel in the cintrol room.

b. Atmospheric Radiation Monttoring The atmospheric radiation maattoring subsystem is

- comprised af monitors of the fixed and mavable type. Each f txed atmospherte monitor is comprised af a particulate measuring channel, todine measuring channel, and a gaseous measuring channel. The air sample that passes through each of these channels is abtained by means af an isokinette sampler and a pump assembly. Samples are obtained by means )f a sampling head placed in a ventilation duct. The sampling heads are equipped with 2ne ar more tsokinette nozzles such that, with a sample flow )f I cfm, the velocity af the air sampled is the same as the velocity of the air in the duct. Mavable atmospheric monitors 2n carts are typt-cally used in the spent fuel handling area during ref ueltng operations and in the r:diochemical laboratory during laboratory sample preparation aperations. These montests are also supplemented by various portable radiation monitoring equipment. Each monttar c antains three channels for particu-late, todine, and gaseous monttaring, respectively. 7-8 Am. 10 1704 I39

Visual high-radLattan alert /12w-level alarms are pravided f or each channel at the 1 scal monitor sampling station f or all atmospheric monitors. Each channel shares an audtble alarm at the local manttor sampling station.

c. Ltquid Radiation Monttoring The liquid radtattan monitoring subsystem c>n-stats af monitors each of which consists af a sampler, setatillation detector, and Cancrol Ra>m ratemeter module. The monttars indicate in the Contral Room in the individual ratemeter madules and two common recorders. The Primary C))lant Letdown monitor also contatus a high range channel consisting af a GM detectar that monitsrs the same sample but uses a separate Contral R)om ratemeter module.

2. Fire Detection Devices Fire protection at TMI is provided by (1) the Fire Service Water System, (2) the Ralogenated Fire Suppres-ston System, and (3) the CO 2 Fire Dctinguishing System.

Ihe Fire Service Water System is a full-loop, piped system that supplies water for (1) sprinklers, (2) deluge water spray, (3) fire hydrants and (4) hose connections that are located such that they provide f Lre protection for all major areas af the plant and atte. A 100,000 gallon Altitude Tank provides the source of water to maintain system piping full and pressurtzed. ThLs provides a method of monitoring system piping tutegrity and provides an intet.1 supply of water f or suppressing a fire. In the event a fire accurs, and either an automatte or manual system is tuttiated, the Fire Service Water System piping pressure will decrease which will cause, sequentially, sne electric and two diesel fire pumps to start as necessary to meet syster flyv requirements. The Halogenated Fire Suppresstan System is pr2vided Ln the plant ventilating air intake and tunnel to inhlbLt combustion af any fuel /atr atxture whleh might enter the intake structure. Detection af an embryante exploston releases Halon 1301 gas into the mtxture in suffletent quantity t) render Lt Lne)m-bustible within a fraction af the time required f or the explis tar. ta reach destructive proportions. 7-9 Am. 10 1704 140

The CO, Fire Extinguishing System provides fire protecElon for the 338 foot Elevation Relay Room. A supply of low pressure CO is maintained in a refrigerated storage unit.2 Se CO is discharged 2 into the Relay room, after a time delay, when a manual pushbutton is depressed or when a thermostat indicates high room temperature.

a. The above systems are actuated either manually or automatically in response to signals from detectors monitoring conditions in the protected area. The following is a list of detectors used in various systems discussed above:

(1) Thermostats Thermostats are used to monitor the tempera-ture in the protected area. khen the temperature in the area exceeds the setpoint of the thermostat, a signal is either sent to an automatic control circuit which will function to suppress the fire or sound an alarm causing an operator to activate a suppression system, or both. (2) Embryonic Explosion Detector he embryonic explosion detectors consist of a photocell and a pressure detector. The photocell will sense the initial flash of a flame front, and the pressure detector will sense the pressure wave as the flame propo-gates. h is detector will react to an explosion before the explosion has time to reach destructive proportions. This signal is sent to the Halogenated Fire Suppression System, and will alarm in the Control Room. (3) Temperature Rise Detectors P Temperature rise detectors will monitor the protected area, and send a signal to either an automatic control circuit or an alarm in the Control Room, or both, if the rise in temperature reaches a setpoint. (4) Fusible Link

• l Fusible links are used in sprinkler systems to hold the sprinkler heads closed. k'h e n the temperature at the sprinkler head rises to a l 7-10 Am. 10 1704 141

I preset temperature, the link melts releasing the sprinkler head which then opens. In the case of a wet sprinkler system, water is then discharged through the sprinkler to suppress the fire. In the case of a dry sprinkler system, air is bled out through the sprinkler head, which depres-surizes the sprinkler header and causes a water supply valve to open thereby supplying water to the sprinkler. (5) Flow Detectors Flow detectors are used in the sprinkler syctems (refer to (4) above), to sense the flow of water through a header supplying a sprinkler system, there-by indicating that the sprinkler system has been actuated. This flow detector will send a signal to an alarm in the Control Room to alert the operator as to the status of that sprinkler system. (6) Smoke and Vapor Detectors Ionization type smoke and products of combustion detectors are installed in the Control Building air return ducts. When smoke or products of combustion are detected, an a? arm will sound in the Control Room and the ventilation syste.n will automatically shutdown. Combustible vapor detectors are used to monitor Auxiliary Building plant ventilation exhaust ducts.

3. S_eismic Monitoring Strong motion recording systems at the TMI Nuclear Station measure ground motion and structural vibra-ting response caused by an earthquake occurring in the vicinity of the site. A cassette magnetic tape recorder located in the Unit 1 Control Room receives information supplied by the triaxial sensor units which are firmly mounted on the Re?ctor Building.

One triaxial sensor unit is attached immediately outside of the containment wall at the base of the Reactor Building. A second triaxial sensor is . situated along the same Reactor Building axis, but is at tached to the Reactor Building ring girder. The triaxial sensor units begin to supply seismic data to the magnetic tape recorder after a signal is sent to

• the sensors by a remote st arter unit. A remote starter unit attached to the base of the Reactor Building provides a signal for it s systems sensor units when the starter unit detects a ground accelera-tion greater than a preset threshold level. The remote starter also actuates an annunciator labeled 7-11 Am. 10 1704 142

o Capitol City Airport o Harrisburg International Airpert o State Turnpike Authority - Harrisburg The NWS will also institute emergency baloon runs to collect data upon request. Air stability determinations are also provided, with information received from weather stations in Pittsburgh; Washington D.C. ; Bing-hampton, NY; and Atlantic City, NJ, to the TMI Nuclear Station upon request.

3. Department of Energy (DOE) Radiolorical Assistance Proeram The DOE has established several regional coordinating offices under the Radiation Assistance Program (RAP).

When notified of an emergency, the RAP Team will respond with equipment and personnel as required to assist in the performance of protective actions. The equipment available consists of, but is not limited to the following: o Portable radiation survey instrumentation o Portable personnel monitoring equipment o Mobile Laboratory facility o In addition, laboratory facilites at the Knolls Atomic Power Laboratory and Pittsburgh Naval Reactors arc available for RAP use, and will be transported to the area for use if necessary. 4 Bureau of Radiation Protection (BRP) (a) The BRP maintains laboratory facilities in Harris-burg consisting of the following instruments or their equivalent: o Gamma Spectroscopy - ND-4420 with NaI and Geli detectors. o Proportional Counters - (1) Bechman Wide Beta, (2) Gamma products and (3) NMC-PC-4 o Liquid Scintillation Bechman LS-133 o Thermoluminescence Losimetry - Panasonic UD-702D Reader The following additional instruments, or their equi-valent, are expected to be purchased in the near future: o Camma Spectroscopy - ND-6620 (or equivalent) with 5 Infrinsic Germanium detectces o Thermoluminescence Dosimetry a second Panasonic UD-702D Reader (b) The BRP also maintains the following portable instruments and equipment, or their equive!xat, for l field use: o Routine / emergency mobile laboratory with analyzer. 7-15 Am. 10 1704 143

o Routine monitoring van used for collecting TLD's in the field, o Eberline SAM-2 o Beta gamma survey instruments o Air samples o Respiratory Protection (c) Environmental monitoring is performed by the BRP as follows: (1) TLD monitoring is performed at twelve (12) locations in the area of TMI. (Additional stations are planned for future install-ation.) These TLD's are routinely read either monthly or quarterly, depending on location; the frequency is increased during an emergency situation in which a radiological release is involved. (2) Water samples are collected at three (3) locations on the Susquehanna River and analyzed on a monthly basis. (3) Milk samples are collected at two (2) local dairies and analyzed on a quarterly basis. (4) Air samples are collected at two (2) locations and analyzed in a weekly basis. (5) Fish and vegetable samples are collected and analyzed annually (July). 4.7.7 Protective Facilities and Eculpment Personnel protective action at TMI is a function of the nature of the hazards, (i.e., preparing for a hurricane is somewhat different from preparing for radiological hazards). Preplanned responses to the basic hazards, high wind, flooding, earthquakes, and radiation exposure are an integral part of the TMI Emergency Plan and are detailed in other sections. A fundamental concept in personnel protection is the immediate release and removal of all individuals not essential to the operation, safety, security, and damage control of the plant. Obviously some hazards can occur before any significant protective action can be applied; an earthquake for example. When the situation permits positive action, the appropriate alarms are sounded and all personnel on the site either assume their assigned emergency responsibilities or are assembled at the designated points for accountability prior to release or assignment to an emergency team. The Met-Ed Unit 1 Warehouse is the normal assembly point for all f non-essential personnel in Unit 1. This structure is a pre-engineered metal building with a conventional ventilation system. It is normally manned at all times. Respirators, protective clothing, and most other prottetive equipment for the plant are stored in this warehouse. If required, personnel assembled at this point could be issued protective equipment from stored supplies. 7-16 1704 144 Am. 10

Warehouse No. 2, located west of the Unit 2 Turbine Building is the normal assembly area for all non-essential personnel in Unit 2. Protective facilities include the TMI Unit 1 Control Room. This area is located in seismically rated structures and have adequate shielding to permit safe occupation for extended periods of time without exceeding an exposure limit of 3 Rem. The Control Room ventilation system has redundant fans and chillers and is provided with radiation and smoke detectors with appropriate alarms and interlocks. Provisions have been made for the Control Room air to be recirculated through high efficiency particulate (HEPA) and activiated charcoal filters. Fresh air is drawn through an underground ventilation tunnel which has been provided with protection against combustible vapors, incipient explosions or fires. The tunnel is seismic class I and is also designed for the hypothetical aircraft incident. Scott air packs and respirators are located in the Control Room to permit continued occupancy if ventilation systems fail. An extensive medical aid facility is located below the Control Room in the Health Physics Lab. 4.7.8 First Aid and Medical Facilities First aid facilities at TMI are designed to support a wide range of immediate care ranging from simple first aid up to and including procedures requiring a physician. The most readily available first aid is provided by small kits placed throughout the plant. These kits contain 16 items (units) typically needed to care for minor injuries. The next level of first aid equipment is found at twelve first aid stations located in the following areas: o Unit 1 Circulating Pump House o Training Department Hallway o Unit 1 Health Physics area hallway o Unit 1 River Water Pump House o Industrial Waste Building o Unit 2 Turbine Building near the roll-up door o Unit 2 Control Room o Unit 2 Reactor Building entrance o Unit 2 Auxiliary Building, elevation 305 feet, by the elevator o Unit 2 Circulating Water Pump House o Warehouse No. I north end at the receiving dock o Hallway in the Unit 1 Control Building These stations are equipped with sufficient quantities of the following supplies to serve the expected needs of 100 employees: o Stretcher o Inflatable Arm Splint o Inflatable Leg Splint o Utility Blanket o Large Plastic Sheet o 100 Employee Size First Aid Kits (contents lised in Table 14) . Four specific areas in the TMI Nuclear Station have been set aside as firs t aid facilities. In a medical emergency the particular facility used would depend upon (1) its proximity to the injured party, (2) 7-17

                                                                        ^=

f704 145

480 MAINTAINING EMERGENCY PREPAREDNESS Metropolttan Edisan C >mpany will maintain, as tua separate d acuments, this Emergency Plan and the Emergency Plan Implementing Document. It ts latended that this Plan, alth2 ugh canstdered as part of the Three Mlle Island Nuclear Station Unte 1 Final Safety Analysts Report (FSAR), vill be maintained as a separate document as suggested by the guidance pravided in referene,e 4.10 4. Eff arts will be made es assure continuous emergency preparedness and )perattanal readiness among Met-Ed personnel and the offstte response agencies and organizatt ans. The Vice President Met-Ed has been assigned yverall responsiblitty for emergency planning related to the TMI Nuclear Station. This responsibtitty includes not saly the TMI Emergency Plan and Implementing Document but als) includes its interrelationship with State, federal, and county plans; agreement letters; carporatc policy and plans; and other related plans, programs, and procedures. To assist the Vice President Met-Ed in meeting his assigned responsiblitties, an Emergency Planning Csardinator has been des ig nated. The specif te responsibilities delegated to the individual assigned as the Emergency Planning Coordinator are described in the f ollowing subsecttons and in particular, subsection 4 8 1 3. 481 Org anizational Preparedness 48.11 Training All personnel at the Three Mile Island Nuclear Station vill take part in a formal training program under the direction af the Manager of Training. In general, this l training program pr2vides for the indoctrination of Met-Ed/GPU employees and cantractors in additt2n to prsviding specialized training for Itcensed iperatsrs, health phys tes/ radiation protection personnel, and personnel assigned specific responsiblitties in the emergency arg anization. The Manager af Tralning is responsible for ensuring that personnel La each department receive the appropriate training. He may delegate specialty training responst-blittles :-) persannel qualtfled to perform such training. The training pragram for the TMI Nuclear Statt)n with regards t) the TMI Emergency Plan vill include the f allowing : 1 All Three Mile Island staf f personnel, excep: personnel in the Operations Department, are required t2 attend the General Employee Training Program at least 2nce per calendar year. In addition, the prampt indoctrin-atton af new employees and c>ntractar personnel ts pr7vided f or Ln the Health Phys tes Tralning Pr > gram whleh they are required ta attend prior to receiv Eng the pr Lv Llege af unescarted access 2nstte. With 8-1 Am. 10 1704 146

specific training on dose calculations /proj ection=, pratective acclan guides, and repartable inf orma-tt an will also be provided.

c. The State P) lice will, on at least an annual basis, be invited to participate in a training pr agram that will include a review of the appitcable parts af the TMI Emergency Plan and Implementing Document with emphasis an the classificattan af emergencies, communications, and specific areas )f respanstbtitty.

4. Met-Ed will also provide orientation and training to

1) cal services support organizations as spectfled in respective letters .2f agreement and as required ta ensure a high state af emergency preparedness and response capabtitt; between these srganizations and the TMI Nuclear Station emergency organizat Lon.

The local services support organizations and personnel who may provide 2nstte emergency assistance will be encouraged to become familiar with the TMI Nuclear Station (including the physical plant layout) and key station personnel, and will be invited to attend Emergency Plan orientation and training courses conducted by or for Met-Ed It is anticipated that such training will be provided on at least an annual basis and will be made available to the apprspriate personnel of the following organizations and certain spectfled individuals.

a. Middletown Police Department The Middletown Police Department will be invited to participate in a training program that will include a review sf the applicable parts af the TMI Emergency Plan and Implementing Document with emphasts an the classification af emergencies, g

communications, and specif te areas af respon-sibility.

b. Fire Compantes The local fire and rescue compantes will be invited to partletpate in a training progran that will, as a mintmum, include the following toples: -

(1) Interf ace with the Site Security Force during emergencies. (2) Basic health phys tes indoctrinat tan and train ing . 8-4 Am. 10 1704 147

and agencies will be canducted. The scape af the exercise will test as much af the emerg ency plans (i.e. , Met-Ed/TMI, State, and counties) as is reasonably achievable. The degree )f pubite participatton in this exercise shall be determined by the appro-priate State agencies. 4813 Emerg ency Planning Coordinatar A member of the TMI Nuclear Station staff will be designated as the Emergency Planning Coordinator. His respons Lbtitties shall include, but not necessarily be limited to:

1. Ensuring the c>2rdination af the TMI Emergency Plan with the:

a. State plans (reference 4 10 22 and 4 10 23)

b. Caunty plans (reference 4 10 24 through 4 10 28)

c. TMI Security Plan

d. Met-Ed Emergency Public Inf ormation Plan

2. Ensuring that the information, data, and pracedures detalled in the Emergency Plan Implementing Dacument are consistent with the evidence provided in the TMI Emergency Plan.

3 Ensuring that the Emergency Plan Implementing Procedures are coordinated and interf ace properly with )ther procedures (e.g. Administrative Pracedures, Security Procedures, Health Phystes Procedures, and Training Procedures , etc. ) .

4. Assisting the Manager -af Training in esordinating and/or providing emergency planning related specialty tra lning .

5 Coordinating emergency planning related drills and exercises as described in subsection 4 8 1 2 above.

6. C)ordinating the review and updating of the TMI Emergency Plan and Implementing Document as described in Section 4.8 2 below.

7 Ensuring the matutenance and inventory 2f emergency equipment and suppites as described in Sect Lan 4 8 3 bel >w.

8. Maintaining himself current with respect ta changes in f ederal regulattans and guidance. that impact emergency planning activtties.

8-9 Am. 10 1704 148

7.3.5.2 Sample Drains The TM1-1 Nuclear Sampling sample and analysis drains are routed to the TMI-1 Auxiliary Building Sump to be processed by the Liquid Radwaste System. In the event of a plant accident accident water including reactor coolant would be discharged to the sump as the result of sampling and analysis. Since the sump is not sealed radioactivity from the accident would escape uncontrollably into the Auxiliary building from the sump. To prevent this from happening, modifications will be made to pipe the radiochemical laboratory drains to the Miscellaneous Waste Storage Tank either directly or by way of an intermediate collection tank and pump (s). The Miscellaneous Waste Storage Tank would contain the laboratory wastes because the tanks gas space is connected to the vent header. Intermediate tanks and pumps that may be used would also be vented to the waste gas system. The laboratory waste collection modification. will be operational by October 1, 1980. 7.3.5.3 Improved In *lant Radioiodine Monitoring Instrumentation Both the TMI-1 Control Room and the Health Physics Laboratory will be provided with analytical equipment capable of distinguishing radioactive Iodine-131 from other airborne radionuclides. The in-strumentation will consist of NaI crystals with a single or dual channel analyzer capable of setting low-level discrimination and window width. Iodine-131 settings will be pre-establish as with other nuclides such as Rb-88 and Cs-137 to allow for short-term identification of radiological hazards and determination of the need for respiratory protection. The equipment will be checked quarterly and recalibrated as neces sa ry. Functional checks will be performed monthly to insure instrument operability. 7.4 Affect of TMI-2 Recovery on THI-1 Operation Activities in TMI-1 related to radwaste processing and activities in the common fuel handling building will not be affected by the THI-2 recovery program. As demonstrat.ed in Section 7.2.3, TMI-I does not have to rely on any TMI-2 facilities for the processing of radwaste. Section 7.2.1 describes specifics to be taken to isolate the radwaste piping systems of the two units. Through the isolation of the piping system, interface between the two unit's radwaste systems will be eliminated. Waste processing activities related to TMI-2 will be performed in the fuel handling building during the recovery program. These activities will not affect activities in the auxiliary building because the areas will be separated by an environmental barrier (Section 7.2.2). Communi-cation of the air spaces (TMI-1/TMI-2) of the fuel handling building will be minimized with appropriate modifications of the ventilation equipment in the building. Continuous access to the fuel handling building is not required for the safe operation of TMI-1 (with environmental barrier installed). 7-15 Am. 10 1704 149

APPENDIX 8A Introduction As discussed in response to Question 39, Supplement 2, Part 2, GPU has embarked on an analysis of the TMI-1 plant response using RETRAN/GPU-01. This code is a modification of the RETRAN (see Reference 24) one-dimensional thermal hydraulic analysis code developed for the Electric Power Research Institute. Model Description Two basic models have been developed for TMI-1. The first is the one-loop model shown on Figure 8A-1. This nodali:ation scheme has been developed to provide relatively detailed analysis results for cases when non-symetric secondary effects are not important. The second model is shown on Figure 8A-2. This model provides a representation of the RCS and secondary system as a two loop model. A nodali:ation of this type allows the modeling of non-symetric effects in either the RCS or secondary systems. The control systems for both models are the same,with the exception that secondary system controls are modeled separately for each secondary system. RETRAN/GPU-01 models all of the reactor protective system and SFAS trips and initiating signals. The model also initates SFAS on loss of all four reactor coolant pumps and 20% voiding at the pump inlet. Secondary system pressure control is explicitly modeled to separate the effects of the turbine bypass, atmospheric exhaust, and safety valves. Feedwater and emergency feedwater are terminated upon initiation of a feedwater latch signal (600 psig in the steam generator). The emergency feedwater system is modeled to separate the motor driven and steam driven pumps, with diesel generator load sequencing and mechanical flow coast up accounted for in each system, separately. Emergency feedwater is initiated by the RETRAN trips when an initiating condition is sensed. OTSG 1evel is controlled at a low or high level, depending upon the availability of the RCS pumps. The HPI/ makeup system is also modeled explicitly. The normal makeup pump controls pressuri:er level at the normal set point of 220 inches via the makeup control valve (MU-V17). Upon SFAS initiation, normal makeup is isclated and HPI is initiated, with flow varying with RCS system backpressure. Proposed Analyses Table 8A-1 lists the analyses which we intend to undertake using RETRAN/GPU-01. There are four basic accidents / transients of interest: loss of feedwater, loss of offsite power, station blackout, and feedwater line break. We may also perform steam line break analyses if a suitable model can be developed. None of the above analyses are intended for use in support of the TMI-1 restart effort. However, other analyses may be performed as docket analyses in the future. 8A-1 Am. 10 1704 150

The scope of analyses is intended to provide a more detailed understanding of the overall plant response to a broad spectrum of likely design basis transients. Analysis Results To date, several analyses have been completed. Results are discussed briefly below. Figure 8A-5 provides results of the base case loss of feedwater transient. Table SA-5 lists key analysis assumptions. No equipment failures are assumed and no operator action is taken. The reactor trips as a result of the loss of feedwater trip. The analysis was terminated after 10 minutes. Hot and cold leg temperatures have converged, and pressuri:er level is being restored by the normal makeup pump. Pressuri:er spray is actuated very briefly, while the RCS pressure never approaches the PORY setpoint. Emergency feedwater initiates as a result of the loss of feedwater pumps. Both the motor driven and the steam driven pumps achieve full flow, and flow is automatically controlled to maintain OTSG 1evel at 30 inches in the startup range. Secondary system pressure is limited by the safety valves, atmospheric exhaust valves, and turbine bypsss in the first 50 seconds, and by the turbine bypass valves thereafter. Figure 8A-6 provides results of a loss of feedwater event with an EFW system which has flow limiting devices. Table 8A-6 provides key analysis assumptions. Two motor driven pumps run, but flow is limited to 740 gpm. The analysis was carried out until the plant achieved stable conditions: RCS temperature and pressure were constant, pressurizer level was restored, the secondary side heat sink was provided via the turbine bypass system. As can be seen from the "PSAT meter graph", the subcooled condition of the RCS was never challanged. Figure SA-7 provides results of the base case station blackout. Table SA-7 lists key analysis assumptions. After one hour, RCS conditions have still not converged, i.e. , hot leg temperature and RCS pressure are still decreasing and pressurizer level is about to go off scale (although the pressurizer would not be empty). As in the other analyses, no operator actions were assumed, and no additional failures were imposed. The analysis does provide some indication of the time frame for operator actions to restore onsite AC power. 8A-2 Am. 10 1704 151

TABLE 8A-1 PROPOSED RETRAN/GPU-01 ANALYSES OF B11-1

1. Loss of Offsite Power (LOOP)

Case .; Base Show plant response to LOOP and transition to natural circulation. Case 2: 1840 gpm EFW and 1% Examine overcooling potential with l decay heat (max cooldown case) 200% EFW and minimum decay heat. Ca.=e 3: Stuck open OTSG safety Examine plant performance for first valves (17'. of design flow) 10 minutes of secondary side depressurization using one OTSG model. Case 4: 100'. EFW and flow limitation Examine long-term plant response with EFW flow limitation and LOOP from 100% power. Case 4a: 1004 FFW and flow limitation Same as case 4, but evaluate importance of modeling pressurizer spray bypas, flow on correct prediction of pressurizer pressure. Case 5: l'. decay heat with 200'. EFW Examine effect of flow limiters in flow limitation EFW system on max cocidown case. Case 6: 2 OTSG model with stuck Evaluate effect of non-symetric open r-lief valves cooldown on secondary side. Case 7: 2 OTSG model 6 no EFW Evaluate non-symetric loss of heat to 1 OTSG sink. Case 8: EFW in superheat region Evaluate effect on transition to natural circulation of putting EFW into superheat region rather than downcomer. Shows more realistic plant response. Case 9: No EFW Evaluate plant response to loss of heat s. < k undar natural circulation conditions. II. Station Blackout Case 1: Base Look at long-term plant response to event, including voiding in the RCS LOOP. Case 2: I gpm pressuri:cr leakage Effect on plant response due to cool-and min. decay heat down of pressurizer steam space. 1704 152

TABLE SA-1 (Cont . ) Case 3: Raise atmospheric exhaust Evaluate efficacy of blackout setpoint proc dure, which calls for this action. III. Loss of Feedwat er l Case 1: Base Show plant response to LOFK with - no equipment failures and no operator actions. Case 2: 460 gpm EFW Flow Examine plant response to operation of only one motor-driven EFW pump. Case 3: EFW flow limitation Look at plant transition to stable shutdown with EFW flow limited. Case 4: 1840 gpm Contrast system response against EFW flow limitation case. Case S: Failure of EFW 1evel Evaluate effect of continuous EFW control flow causing overfill of OTSG. Case 6: Trip from low power and Look at plant response below t bine min. decay heat trip threshold. Case 7: Case 2 or TMI-2 in 2 OTSG Compare 1 6 2 LOOP results. model Case 8: Partial LOFW Compare response to complete LOFW 6 evaluate ICS response. Case 9: No EFW Look at time available before HPI must be initiated. Case 10: Turbine stop valve fails Limiting overcooling case following open a LOFW. Case 11: Manual actuation of HPI Look at effect on pressuri:er level of operator initiating a second HPI pump immediately after reactor trip. IV. Feed Line Break Accident Case 1: EFW initiation on feed / steam Demonstrate plant can tolerate the AP design basis feedline break accident. Case 2: EFW initiation on low OTSG Evaluate timeliness of this signal level for feedline break.

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TABLE 8A-1 (Cont ) Case 3: Partial break which causes Determine if operator response is gradual loss of OTSG 1evel as can be reasonably expected, and no feed / steam AP signal V. Steam Line Break Accident Case 1: Benchmark of TMI-2 docket Establish benchmarked code for analysis use on TMI-1. Case 2: TMI-1 design basis analysis Demonstrate means required to establish long-term safe shutdown. Case 3: Break from startup condition Determine if startup from low with flooded no::les OTSG 1evel is necessitated by safety concerns. O 1704 154

TABLE SA-3 LOSS OF OFFSITE POWER CASE 2 MAKEUP: AVAILABLE LETDOUN: ISOLATED AT T=S SEC EFU MECH: AVAILABLE EFU STEAM: AVAILABLE , EFU CAPACITY: 1840 GPM ATMOS DUMP: AVAILABLE SETPOINT=1025 PSIA TURBINE BYPASS: UFAVAILABLE SMALL SAFETIES: 1050/1028 PSIA BANK 1: 1065/1033 PSIA RCS PRESS HTRS: 2 BANKS ( 126 KU ) PRESS SPRAY: UNAVAILABLE RC PUMPS  : COAST DOUN BEGINNING T=0 RPS TRIP  : LOOP RPS TRIPS BYPASSED : NONE SFAS TRIPS BYPASSED : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS  : 2 0FFSITE P0UER: UNAVAILABLE G T=0 DECAY HEAT: 0.1 X ANS PORU: 2450/2400 PSIG SAFETY: 2500/2475 PSIG , 1704 155 Am. 10

TABLE 8A-4 LOOP ANALYSIS ASSUAIPTIONS LOOP CASE 5:1145 GP.\1 EFh' S l'. ANS DECAY HEAT MAKEUP: AVAILABLE LETDOUN ISOLATED AT T=5 SEC EFU MECH: AVAILABLE EFU STEAM: AVAILABLE EFU CAPACITY: 1145 GPM ATMOS DUMP: AVAILABLE SETPOINT=1025 PSIA TURBINE BYPASS: UNAVAILABLE SMALL SAFETIES: 1060/1028 PSIA BANK 1 : 10S5/1033 PSIA BANKS 2&3 : 1070/1038 & 1072/1040 PSIA RCS PRESS HTRS:2 BANKS (126 KU) PRESS SPRAY: UNAVAILABLE RC PUMPS  : C0AST DOUN BEGINNING T=0 RPS TRIP  : LOOP RPS TRIPS BYPASSED : NONE SFAS TRIPS BYPASSED : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS  : 2 0FFSITE POWER: UNAVAILABLE O T=0 DECAY HEAT: 0.1 X ANS PORU: 2450/2400 PSIG SAFETY: 2500/2475 PSIG p 9 1704 156 Am. 10

TABLE 8A-S LOFN BASE CASE ANALYSIS ASSUMPTIONS MAKEUP: AVAILABLE LETDOUN: ISOLATED AT T = 5 sEc EFU MECH: AVAILABLE EFU STEAM: AVAILABLE EFU CAPACITY: 1900 GPM TURBINE BYPASS: AVAILABLE SETPOINT=1025PSIG ATMOSPHERIC DUMP: AVAILABLE SMALL SAFETIES: 3% RESET POINT BANK 1 : 3% RESET BANKS 2&3: 3% RESET PRESS HTRS: 5 BANKS PRESS SPRAY: AVAILABLE 2220/2170 PSIA RC PUMPS  : AVAILABLE RPS TRIP  : TURBINE RPS DEFEATED  : NONE SFAS TRIP 1500 PSIG RCS 9 PSIG CONTAIN PRESS DEFEAT DIESEL GENERATORS  : 2 0FFSITE POUER: AVAILABLE PORU: 2450/2400 PSIG PZR SAFETIES

• BS00/247S PSIG 1704 157 Am. 10

TABLE SA-6 LOFK CASE 2 ANALYSIS ASSU',!PTIONS MAKEUP: AVAILABLE LETDOUN: ISOLATED AT T = 5 sac EFU HECH: AVAILABLE EFU STEAM: AVAILABLE EFU CAPACITY: 700 GPH 100%-FLOU LIMITER TURBINE BYPASS: AVAILABLE SETPOINT=1025PSIG ATMOSPHERIC DUMP: AVAILABLE SMALL SAFETIES: 3% RESET POINT BANK 1 : 3% RESET BANKS 2&3: 3% RESET PRESS HTRS: 5 BANKS PRESS SPRAY: AVAILABLE 2220/2170 PSIA RC PUMPS  : AVAILABLE RPS TRIP  : TURBINE RPS TRIPS DEFEATED : NONE SFAS TRIP STATUS  : 1S00 PSIG RCS

4 PSIG BLDG PRESS DEFEATED DIESEL GENERATORS  : 2 0FFSITE POUER: AVAILABLE DECAY HEAT: 1.0 ANS PORV: 24SO/2400 PSIG -

PZR SAFETIES : 2S00/247S PSIG. Am. 10 1704 158

TABLE SA-7 CEE 1 STATION BLACKOUT ANALYSIS ASStf!PTIONS MAKEUP: UNAVAILABLE - LETDOUN ISOLATED AT T = 5 SF.C EFU MECH: UNAVAILABLE EFU STEAM: AVALIABLE EFU CAPACITY: 920 GPM ATMOS DUMP: AVAILABLE TURBINE BVPASS: UNAVAILABLE SMALL SAFETIES: 3% RESET POINT BANK 1 2 3% RESET BANKS 2&3: 3% RESET PRESS HTRS: UNAVAILABLE PRESS SPRAY: UNAVAII.ABLE RC PUMPS  : UNAUAILABLE RPS TRIP  : LOOP DIESEL GENERATORS 0 0FFSITE POUER: UNAVAILABLE DECAY HEAT: 1.0 X ANS PORU: 24SO/2400 PSIG - PRESSURI2ER SAFETY: ES00/247S PSIG Am. 10 1704 159

TABLE SA-8 LOOP CASE 4 ANALYSIS ASSUMPTIONS MAKEUP: AVAILABLE LETDOUN: ISOLATED AT T=5 SEC EFU MECH: AVAILABLE EFU STEAM: AVAILABLE . EFU CAPACITY: 490 GPM ATMOS DUMP: AVAILABLE SETPOINT=iO25 PSIA TURBINE BYPASS: UNAVAILABLE SMALL SAFETIES: 1060/1028 PSIA BANK 1 : 1065/1033 PSIA ~ BANKS 2&3 : 1070/1038 & 1072/1040 PSIA RCS PRESS HTRS:2 BANKS (126 KU) PRESS SPRAY: UNAVAILABLE RC PUMPS C0AST DOUN BEGINNING T=0 RPS TRIP  : LOOP RPS TRIPS BYPASSED : NONE SFAS TRIPS BYPASSED : 4 PSIG CONTAINMENT PRESSURE DIESEL GENERATORS  : 2 0FFSITE POWER: UNAVAILABLE 6 T=0 DECAY HEAT: 1.0 X ANS - PORU: 2450/2400 PSIG SAFETY: 2500/2475 PSIG 1704 160 kn. 10

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10.0 CROSS REFERJNCE TO ORDER RECOSDiENDATIONS 10.1 INTRODUCTIO ; The August 9, 1979 Order and Notice of Hearing issued by the Commission listed numerous actions recommended by the Director of Nuclear Reactor Regulation (NRR). The.ee recommendations are listed in Section 10.2. The section oi this report that covers the recommendations is referenced or a response is given. A number of the reco=mendations require additional guidance or have been auended/ modified during the course of their development especially those related to ISE bulletins 79-05A, 05B,a nd 05C and NUREG 0578. Met-Ed is therefore responding to these recou-mendations as they are currently understood. 10.2 SHORT-TERM RECOSNENDATIONS AND MET-ED RESPONSES Recommendation Response 1(a) Auxiliary Feedwater Upgrading Section 2.1.1.7, S1 P1 Q1-10 and S1 P2 Q1-15 1(b) Auxilary Fee'dwater Operating Sections 3.1.1 and 3.1.4, Procedures also see la above 1(c) Reactor Trip en Section 2.1.1.1 Loss of Turbine /FW 1(d) Complete Analysis for Small Section 3.1 Break LOCA's and Revise Procedures 1(e) Retraining of all Reactor Section 6.0 Operators 2 I6E Bulletins IEB 79-05A Item 1 S1 P1 Q29 Item 2 See Section 10.3.1, and S1 P1 Q34 Item 3 Section 3.1.1 and S1 P1 Q32, and 50 Item 4 Sections 3.1.1 6.2 and S1 P1 Q32, and 52 Item 5 Sections 3.1.2, 3.1.3 and S1 P1 Q53, and 54 Item 6 Section 2.1.1.5 Item 7 Sections 3.1.2 and 3.1.3 10-1 Am. 10 1704 245

Recommendation Response Item 8 Section 11.2.1 & S1 P2 Qll Item 9 Section 2.1.1.5.3, S1 P1 Q59 Item 10 Section 3.1.3, S1 P1 Q60 Item 11 S1, P1, Q61 Item 12 Section 3.I' Table 3.1-1 (AP 1044) LEB 79-05B Items 1 & 2 Sections 3.1.4 and 6.2 S1 P1 Q55, and 56 Item 3 Sections 11.2.3 and 8.1 (3ased on B&W analyses submitted May 7, 1979), 8.2 and S1 P1 Q30 Item 4 Sections 3.1.1 and 2.1.1.1 Item 5 Section 2.1.1.1, and S1 P1 Q31 Item 6 Section 3..? Table 3.1-1 (AP 1044) 51 P1 Q57 Item 7 Section 11.0, and S1 P1 Q38, and 58 IEB 79-05C Item 1 Section 3.1.1 & S1, P1, Q62 Item 2 (Later) Item 3 Section 3.1.1 & S1, P1, Q33 Item 4 Sections 3.1.1 and 6.0 Item 5 See NUREG 0578 Item 2.1.9 below Item 1 (Long Term) Sections 2.1.2.5 and 8.2.4

3. Emergency Plan Upgrading Section 4.0 4 TMI-1/TMI-2 Radwaste Ventilation Section 7.2 and Sampling Separation

5. TMI-l Radwaste Management Capa- Section 7.3 bility

6. Organization and Resources Section 5.0 10 -2 Am. 10 1704 246_

Recommendation Response

7. Financial Qualifications To be Submitted Separately

8. THI-2 Lessons Learned Recom-mendations - NUREC 0578 2.1.1 Section 2.1.1.3 and 51 P1 Qll

                                 & 14 S1 P2 Q18 & 30.

2.1.2 See Section 10.3.2 also see S1 P1 Q16 and S1 P2 Q19 2.1.3.a section 2.1.1.2, S1 P1 Q13 & 15 and S1 P2 Q20, 36 & 37 2.1.3.b Section 2.1.1.6, S1 P1 Q17, 18, 19, 20 & 39 and S1 P2 Q39, 32, 93, 94 6 95 2.1.4 Section 2.1.1.5, S1 P1 Q21-28 2.1.5 Section 2.1.1.4 2.1.6 Section 2.1.1.8 2.1.7 Section 2.1.1.7 & 2.1.2.6 2.1.8a Section 2.1.2.4 2.1.8b Section 2.1.2.7.2 2.1.8c Section 7.3.3.3 l 2.1.9 Sections 3.1.1, 6.0, 8.1 2.2.1.a S1 P1 Q40 and S1 P2 Q25 2.2.1.b Section 5.0, S1 P1 Q42 and S1 ~ P2 Q27 i i 2.2.1.c Section 3.0, S1 P1 Q43, S1 P1 Q28 and Q29 i 2.2.2 Section 4.0 and S1 P1 Q44 g 2.2.3 Not Applicable until NRC Regulations are revised 10-3 Am. 10 1704 247

dropped approximately 219 inches and Tave dropped to 536*F following the trip; both of these related parameters exceeded their normal response to a turbine trip. The apparent cause was the failure of two main steam safety valves, MS-V21A and MS-V205, to reseat. This allowed an initial drop in OTSG header pressure to 950 psi until the turbine bypass valves adjusted to compensate for this additional steam relief. Although this was a deviation from the expected performance it is not considered to be signifi-cant since no Limiting Conditions for Operation were violated, pressurizer level remained on scale, and the turbine bypass valves were more than adequate to control header pressure. Summary of Corrective Action Since the above review did not identify any significant devia-tions from expected performance, no major corrective actions were undertaken. Minor corrective action such a s rechecking of instrument setpoints were performed. Details concerning each transient that was reviewed and the specific rrective actions taken are included in Appendix 10A. Neith'r of these transients were reported as reportable occurrences. 10.3.2 - Performance Testing for PWR Relief and Safety Valves Section 2.1.2 of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Reco=mendations", recommends that licensees,

                       " Commit to provide perfo' mance verification by full scale prototypocal testing for all relief and safety valves. Test conditions shall include two-phase slug flow and sub-cooled liquid flow calculated to occur for design basis transients and accidents."

In response to NRC recommendations on safety and relief valve testing, Met-Ed will participate in the EPRI program to conduct performance testing of PWR relief and safety valves. The overall objective of the EPRI program is to obtain experience ' and insight to assure that safety and relief valve systems can perform as required to prevent overpressurization of the primary coolant boundary. The program is focused on verification of the key operational requirements concernf..g the hydraulic / structural performance of relief valves, safety valves, and discharge piping:

1. That power operated relief valves open and close as required .

2. That safety valves open and close as required; there is a difference between closure and reseating.

3. Flow capacity through the safety valves is sufficient to prevent overpressure of the primary system.

4. The integrity of the discharge piping is maintained.

10-5 Am. 10 1704 248

While there are many relief and saf ety valves in service there are only a few generic types which define the specifics of the test program. Since existing facilities preclude full-scale testing at this time, a two phase program is being undertaken by EPRI:

1. Existing test facilities will be used for performance testing of small safety / relief valves. Testing will occur under steam, water, and appropriate two-phase conditions to ascertain whether safety / relief valves will open, close, and relieve suf ficient fluid to protect the primary system pressure boundary.

2. In parallel with phase 1, facilities that will allow testing of large safety / relief valves will be designed and constructed.

Present schedules call for scoping tests on relief valves which require the minimum in test facilities to be initiated during April, 1980 followed by safety valve tests, and generic safety / relief valve system tests, to be completed by July, 1981. The expanded valve test facility will be in place by July, 1981. Scheduling of test facilities and other uncertainties could result in a longer schedule. Met-Ed believes, however, that substantive test data can be obtained by July 1981. Additional detail concerning the test program is contained in a letter from W. J. Cahill, Jr., to Harold Denton, Regarding: " Program Plan for the Performance Verification of PKR Safety / Relief Valves and Systems, December 13, 1979," forwarding letter undated. 10-6 Am. 10 1704 249

RESPONSE TO APPENDIX A 0F QUESTION 10, SUPPLDIENT 1, PART 1 See response to Question 3, Supplement 1, Part 2 Am. 10 1704 250

SUPPLEMENT 1, PART 1 QUESTION: 27a. (Restart Submittal Table 2.1-2) Valves O!-V1, O!-V2, CM-V3, and CM-V4 do not receive a diverse safety grade automatic isolation signal. This is unacceptable. Modify your design accordingly.

RESPONSE

The subject valves will be provided with a diverse containment isolation signal on 1600 psig ES (high pressure injection actuation signal). Am. 10 1704 251

SUPPLDIENT 1, PART 1 QUESTION: 27d. Valves RB-V2A and RB-V7 are listed with three identical options and accompanying note. The first option provides automatic isolation only on hi-hi centainment pressure. This option is unacceptable. Options two and three are acceptable providing that the hi-containment pressure isolation signal is retained. Modify your description by deleting the multiple choice options and providing the specific details of your design consistent with the above evaluation.

RESPONSE

We concur with your comment and have directed our design effort to isolate penetrations No. 421 and 422 with valves RB-2A and RB-V7 on both a hi-containment pressure , (4 psig) and also on 1600 psig R. C. pressure. Our ' design will also provide for the automatic initiation of the emergency R. B. cooling system when the normal R. B. Cooling is isolated by either of the above signals. Am. 10 1 7 0 4 i) 5 2.

SUPP_L,EMENT 1. PART 2 QUESTION

3. Your response to Question 8 is not complete. Provide the B5W study on transients such as loss of feedwater and loss of offsite power which verify that minimum EFW flow requirements meet the 550 gpm technical specification commitment. Provide the revised TM1-1 Technical Specifications for our review prior to restart. Justify the applicability of the B5W study to TMI-1.

RESPONSE

Attached is the B6W study (document identifier 86-1102587-00) on the auxiliary feedwater flow requirements following a loss of main feedwater. The analyses performed included the following assumptions:

1) The initial power level at the initiation of transient was 2772 Mwt.

2) The reactivity feedback coefficients used were representative of approximately 100 EFPD operation.

3) The ANS 5.1 decay heat curve was used with a 1.0 safety factor. The key input parameters used are documented in Table 3.2-1 of the BSW Report to the NRC dated May 7,1979 and entitled " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant." The input paraceters assumed in this study are applicable for TMI-1 which is only a 2535 Mwt plant. As noted in the B5W study, auxiliary feedwater flow rates as low as

• 370 gpm were found to provide satisfactory performance.

Additional work has also been done by BGW to demonstrate that 500 gpm auxiliary feedwater flow is adequate following upset transients such as the loss of offsite power and the loss of normal feedwater. This analysis is also attached. The Technical Specifications for EFW will be provided in January,1980. The acceptance criteria for the minimum auxiliary flow rate were that (1) the pressurizer does not go solid and (2) the electromatic relief valve does not actuate. An auxiliary flow rate of 500 gpm was found to meet these criteria. The assumptions used for these analysis are conservative for Till Unit #1. Am. 10 1704 253

ATTACH!1ErlT TO QUESTION 3, SUPPLEf1EfR 1, PART 2 AN EVALUATIO:10F A?l E!'.ERGE! CY i FEEDWATER FLOWRATE OF 500 GFil FOR 256812T St.W PLAliTS D r o ogggg Introduction The purpose of this report is to confirm the results of an investigation of a 500 gpm emergency ficurate for the 2558 !*Jt 177 FA units. The transients analyzed, the acceptance criteria chosen, the trar.sient and steady state analysis methods used, and the analysis assumptions will be discussed. Aa general note, the transient analysis carried the event out to a time at which a steady state analysis could demonstrate that the emergency feedwater heat reroval capability exceeded the decay heat produced. Results The results, developed in Reference 8 and previously transmitted in References 1-5, show that a 500 gpm auxiliary feedeater flowrate requirement for the 177 FA plants with power levels of 2568 !Gt or less is adequate to reri.ove decay heat and meet the acceptance. criteria as noted in the next section with regard to loss of feedsater (with anticipatory trip) and loss of offsite power. The adequacy of 500 grr with respect to ECCS requirements is docum.ented in Reference 6. . Acceptance Criteria For the transient analyses, the acceptance criteria chosen were prevention of pressurizer fill and prevention of saturated conditions in the RC hot leg. Steady state analyses were performed on the basis of maintaining subcooled conditions in the primary system. 1704 254

Events Analyzed Loss of feed. vater and loss of offsite power transients were analyzed with the heat removal being performed by one or two steam generators. The fl5S was assumed to be operating with an anticipatory trip system in place which dropped the control rods within five seconds of the loss of feedwater. For the loss of offsite power transient, the reactor trip was assumed at the point where the RC pumps and fece, vater pumps were lost. Assumptions The following input and assumptions were co=on to all the cases (which satisfied the acceptance criteria):

1. Moderator coefficient: ze ro. This is conservative as all the operating plants will have negative moderator coefficients at hot full power.

2. Doppler coefficient. A value of -1.22 x 10-5 ap/ F was assumed.

3. The input primary flowrate was based on 88,000 gpm/ pump. This'is conserva-tive as all operating plants have measured flowrates in excess of this design value; tt s, the initial system al is larger than the actual.

4. The pilot operated relief valve (p0RV) setpoint was 2450 psig.

5. The mergency feed flowrate assumed was 500 gpm. This value was chosen because it corresponded to the minimum flowrate possible at S&W units as of May of 1979 under single failure conditions.

6. The emergency feedwatcr was assumed to reach the steam generator at 40 seconos after the onset of the upset event. This value was chosen because of its conservatism with respect to the 30-second time provided for the first B&W .

unit analyzed.

                     .                                                             704 255

S 5

7. No simulation of reactor coolant makeup or letdown was made because of the inability of the transient analysis methods available to model these phenomena.

8. The initial pressurizer was 220 inches of a 400-inch range.

9. A conservative ccndition of minimum allowable shutdown margin (-15 J.k/L) was assumed. This margin is the minimum su'bcritical margin at HZP by the control rod / fuel loading patterns. BOL shutdown margins are typically much greater than this value.

10. A safety factor of 1.2 was applied to the ANS 5.1 decay heat curve.

11. The initial power level was 102% of 2558 KAt.

12. No pump heat was accounted for in the LOFW transient analysis due to limitations in the CADDS code. Its effect on pressurizer level was estimated. .

13. Final computer runs did not account for spray and pressurizer heater actuation.

LOFW Analysis The results for the LOPJ event ir.dicats t':: tn 3::e;. .cnce criteria wcre met by 500 gpm for auxiliary feed to two or one OTSGs. A total delay of five seconds between tne loss of flow and the CRCM release was assumed. The PORY did not actuate for thcse cases. At the time of the analysis, no transient analyses methods were available to nodel the asymetric case of heat removal with one OTSG. The CADDS';ATBYP version of the CADDS code, developed during the period following the TMI-2 incident, has an option which directs all of the primary flow to o;.. 0:S;. 1704 256

_ . m~- - _ _

                               .                _       _ _     ~.        -

m This model . utilized the 52-node steam generator. The case was modeled by dividing the primary system volume and initial power level in half, doubling the factor on the decay heat curve, and directing all the primary flow to one OTSG. The model does, therefore, require that all heat removal be accomplished by one OTSG. The two OTSG cases were modeled with the two-loop option which directs half the vessel flow to one steam generator. The normal . CADDS version was used only for the two OTSG LOFW case. The results for the one OTSG LOFW heat removal case indicated that the pressurizer level peaked at 231 inches at eight seconds and then fell to a minimum of 138 inches at 55 seconds. The level then increased slowly until it peaked at 171 inches at five minutes. When the transient simulation was terminated at 400 seconds, the level had fallen to 168 inches. After an initial peak of 2355 psia at eight seconds, RC pressure dropped and subsequently reached a peak below 2200 psia at five minutes. It is to be expected that the minimum level encountered in these cases would be significantly above that typically observed in the actual LO?d incicents at the operating plants because the conditions modeled were those that would minimize the heat removal capaoility. The minimum level for the two OTSG case wcs 92 inches. The two OTSG case was terminated at 250 seconds with a level of 125 inches at that point. System pressure remained in the 1900-2200 psia range. Typical results were that the pressure reached a minimum at approximately one minute and then slowly increased until heat removal exceeded heat generated. At that point, system pressure and pressurizer level began to decrease. J ] 1704 257

Steady state calculations were perfor ed to deter--ine the emergency feedaater flowrate requirement as a function of time to remove decay heat. No super-heating of the feedwater was assumed and a factor of 1.2 was applied to the ANS 5.1 decay heat curve. The calculation amounted to a simple heat balance. For an assumed emergency feedaater temperature of 100 F, the point at which tne feedwater heat removal capability matched the decay heat generation was approximately six minutes af ter trip. That is, at six minutes the emergency feeddater flowrate times its enthalpy change from 1000 F to saturated conditions in the generator matched the decay heat level. LOOP Analysis J , Both steady state and transient analyses were performed on the loss of offsite power transients. First the transient analysis will be discussed. The CADDSNATBYP code was used to simulate the LOOP 6Yents. A reactor coolant ficw coastdcan obtained from plant data was input (Refererice 7). At three minutes the code switched from this input flow to a flow calculated from the driving heads available for natural circulation. The results were characteri:ed by the inability of the code to model the correct injection point of the emergency feedwater. The resulting low thern.al center associated with bottom injection prevented accurate modeling of long-term natural circulation. However, some observations were made concerning pressuri:er level. 6r ve two OTSG case, the level was 252 inches when the simulation terminated at 15 minutes. For the one OTSG case, the level was approximately 260 inches at eight minutes. (These one OTSG results for nominal decay heat and 100; power.) As noted earlier, on the basis of heat removal capability, 500 gpm matches decay heat at approximately six minutes. Inus , 5C; gp:" srcui d be ...wa . t: prevent pressurizer fill for the LOOP case. 1704 258 .

                                                      - - -   -.. . _-.n           . :_ . ~

D

                                                     'D { { d N A$'aTDWKR s

The present analysis was not intended to be a natural circulation analysis. Conservative hand calculations were perfomed to assess the possibilities of natural circulation heat removal with one OTSG. A thermal center 20 f t. above the tube sheet of the OTSG was chosen. The thermai center is cefined as the distance above the tubesheet at which the primary coolant ray be considered to change from T Hot to T Col d for primary side density profile purposes . It should be noted that upper injection in tne steam generator favorably impacts the thermal center. A power level of 3.7 percent, which corresponds to the decay heat generated at 21/2 minutes was chosen. The result was that subcooled RCS conditions were maintained by this level for this high power level with RC pressures of only 2030 psia. This is significant because it indicates the high power levels at which a 20-ft. thermal center is adequate. In reality, it is not expected that equilibrium will be achieved in 21/2 minutes in terms of primary' flow or secondary water level. Reactor coolant pressure will in all probability be below 2000 psia at 21/2 minutes. However, the RC purrp coastdown is stili in progress a:J tw transient results indicate that the natural circulation mode may not be achieved for at least 10 minutes. When steady state conditions are achieved, the power level will be well belos 3.7% and the corresponding steam generator water level may be below 20 ft. Thus, the conservative 21/2 minute s tc:.dy state calculation shows that natural circulation on one generator is fenible. Since the conservative transient analysis carried the event past the point where hand calculations confirmed the adequacy of 500 gpm, the analysis was considered satisfactory, subject to the limitations discussed below. 1704.259 m

D**D 'D'W "2 Clarification of Limitations The transient analysis methods for this analysis did not permit explicit modeling of several phenomena. This section will contain those transient analysis limitations and the rationale for justifying the 500 gpm flowrate despite these limitations. The behavior of the emergency feedwater system with respect to establishing and controlling the steam generator water level was not modeled. Constant steam pressure (1030 psia after initial fluctuation) and constant feedwater flow were assumed. For an overheating case, it is expected that feedwater would be fed at a maximam rate, i.e. , feedwater flow would not be throttled during a limiting overneating event. The steam pressure assumed provides a high secondary sink temperature and thus maintains a

                                                                                     ~

relatively low primary to secondary temperature difference. As noted earlier, the transient code for the LOOP case, CADDSNATBYP, models the bottom injection of feedwater (Integral Economizer Once Through Steam Generator - IE0TSG). The 177 FA B&W plants have Once Through Steam Generator: (OTSGs) which provide upper injection of er ersency feec ater. The effect of upper injection vs. icuer iNcction is important in terms of establishing tha characteristics of natural circulatien. The heat removal capability, however, is not dependent on the point of . injection but on the inlet and outlet condi-tions of the secondary side of the generator. Since hand calculations confirmed the feasibility of heat renoval with one OTSG, and steady state heat balance calculations indicate that 500 gpm removes the decay heat level present at six minutes, the heat removal adequacy of 500 gpm is confirmed by achieving acceptable results for 10 minutes for a LOOP with one OTSG removing heat.

• s -

1704 260

                                  , _ ,      ,f . . .m ., . .           .,
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Hand calculations were used to verify that the addition of pump heat would not fill the p'ressurizer for the LORI event. The heat capacity of the RCS was taker: into account for this calculation. However, it may lead to a PORV actuation. Finally, it should be noted that flowrotes less than 500 gpm were not analyzed. It is possible that an analysis could produce acceptable results under the assumptions and conditions outlined herein for flowrates less than 500 gpm. e o e O e 9e e 4 1704 261

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i, Refe rences  ; i

1. R. R. Lange to R. P. Williamson, " Minimum Emergency Feedwater Flow Rate i Requi remen ts ," Duke , 5-9-79. *

2. R. R. Lange to R. P. Williamson, " Minimum Emergency Feedwater Flcwrcte Requi rements," T3.4, 5-25-79. ,.

I

3. R. R. Lange to R. B. Davis /F. J. Levandoski, "Minimua Emergency Feed. vater -

Flowrate Requirements," Toledo T3.4, 6-1-79.

4. C. W. Tally to L. G. Weatherford, " Auxiliary Feedwater Flow Requirements,"  :

Florida Power & Light, T3.17, 6-6-79. i

5. R. R. Lange to L. G. Weatherford, " Auxiliary Feedwater Flowrate Require-ments " AP&L, T3.17.2, 6-25-79. ,

I I

6. R. C. Jones to R. R. Lange. " Auxiliary Feedwater Flowrates," All 2568 Math  !

177 FA Plants," 6-27-79.

7. R. B. Park to J. T. Willse, "RC System Flows for MX-B 177 FA Plants," .

T3.5, 7-5-78. '

8. 32-1102390-00, R. R. Lange.

O e e 1 1704 262

SUPPLEMENT 1, PART 2 QUESTIO"

6. Provide design drawings for the modifications which provide for control room annunciation of all automatic start conditions of the EFW system.

RESPO: The condition which results in an EFW auto start, namely loss of F.W. pump or loss of Reactor Coolant Pumps, are annunciated individually as shown on elementary wiring diagram SS-208-110, S5-208-111, SS-208-112, SS-208-113, SS-209-060 and SS-209-061, provided separately. Operating procedures are relied upon to verify that emergency feedwater is established should any of these conditions exist. The design of the safety grade auto start of the E.F. system incorporates an alarm dedicated to annunciation of automatic actuation of EFW. Details for this alarm feature will be submitted at a later date. 1704 263 Am. 10

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SUPPLEMENT 1, PART 2 OUESTION

32. Provide the electrical elementary drawings associated with the containment isolation modifications so that we may complete our review of this aspect of your design.

RESPONSE

The electrical elementary drawings associated with the containment isolation modifications will be submitted under separate cover by the end of January, 1980. l Am. 10 1704 270

SUPPLEMENT 1, PART 2 QUESTION

53. For Solid Radwaste Systems, provide the following:

A. Description B. Capacity C. Process Control Program D. On-Site Storage Facility E. Expected amounts of Solid Wastes Per Year.

RESPONSE

A. A backfit to the TMI-l Radwaste Solidification System is planned and consists of replacing the existing Protective Packaging, Inc. (PPI) System , which utilizes urea formaldehyde as the solidifying agent with one that uses a Dow Chemical Co. Polymer as the binder. The Dow waste solidification process will be used to solidify radwaste in the form of liquids, slurries, and spent ion-exchange resins. Proprietary technology is used to form stable water-resin emulsions which are then chemically cured to form hard, solid, monoliths. Liquid or slurry waste is stirred with a commercially available binder until a stable vaste-binder emulsion is formed. The mixture is then cured by the addition of two chemicals. The final result is a dispersion of small spherical liquid particles (fine dr.oplets) in a continuous matrix of cured binder. An approved process control program will be incorporated into the system in accordance with Regulatory Guide 1.143. The new system will be physically located in the THI-l Radwaste Solidification room where the PPI System is currently located.' Engineering for this backfit will begin in November 1979 and is scheduled for completion approximately April 1,1980. Equipment installation and checkout should be completed by October 1,1980. B. The permanent solidification system, currently being designed for TMI-l will have the capability of solidifying four 50 Cu. Ft. containers in each eight hour shift. This is equivalent to 120 Cu. Ft. of waste (considering packaging efficiency). This process flow rate can be achieved for the solidification of evaporator bottoms. Used precoat and spent resin process rates are 100 Cu. Ft. (60 Cu. Ft. waste) and 50 Cu. Ft. (30 Cu. Ft. waste) per day respectively. 1704 27i Am. 10

SUPPLEMENT 1, PART 2, QUESTION 53 C. The process control programs (PCP) requested are not available f at this time. The PCP for tne temporary systems being considered are currently being written and are expected to be available by mid January 1980. The design of the permanent system has not been finalized and a PCP cannot be draf ted until the design is co=plete. The PCP will not be finalized until the system is tested using non-radioactive test solutions and resin slurries. System start-up testing will be started in the last quarter of 1980. For any system used to solidif y radwaste, a PCP conforming *o Branch Technical Position ETSB 11-3 (as described in SRPil-4) will be approved by the USNRC prior to operation of the system. D. The solidified waste will be stored until shipment with the Epicore II wastes until a permanent waste storage building is available. The EPICOR-II waste staging f acility will be available for the l temporary storage of EPICOR-1 prefilters and demineralizers. The i combined EPICOR-I and EPICOR-II liner production rate will result in 6 to 7 storage cells being filled each month until the TMI-II auxiliary building water inventory is processed (approximately May 1980). Af ter that time the production rate will be reduced , i to 3 to 4 filled storage cells per month. The current construction I schedule for the completion of storage cells will insure storage capacity in excess of the production rate to the end of 1981. Low activity waste (solidified evaporator bottoms, dry trash and other LSA waste) will not be stored in the EPICOR II storage cells because (1) the radiation levels do not warrant this type of storage, (2) the relatively large volumes (compared to EPICOR II) of waste would rapidly deplete the supply of storage cells and (3) the configu, ration of the low activity waste would result in handling proble s during placement into and retrieval from the cells. Immediate plans for solid radwaste storage, other than EPICOR I, utilize existing space in the TMI-I auxiliary building. The following storage capacities exist:

a. 55 gallon drums unshielded (compacted trash) 200 maximum

b. 50 cubic foot liners (evaporator bottoms unshielded) 8 maximum or 50 cubic foot liners (used precoat unshieldel) 4 maximum plus 50 cubic foot liners (evaporator bottoms unshielded) 4 maximum

c. Spent resin storage - with EPICOR waste This amount of storage would provide storage for up to one month.

Am. 10 1704 272

SUPPLE)!ENT 1, PART 2, QUESTION 53 P The need for additional storage capabilities has been identified and an implementation plan to expand existing capabilities will be completed by February 1, 1980. E. Anticipated amounts of solid radwaste produced per year: 5000 cwft Solidified Evap. Bottoms 3000 cwit Compacted Trash (Dry) 1000 ewit Solidified Resin (Based on a normal operating year with refueling outage). Am. 10 1704 273

SUPPLEMENT 1, PART 3 1704 274

SUPPLEMENT 1, PnRT 3 QUESTION:

1. Your response to questions 26 and 37 does not provide the staff with sufficient information to make an evaluation of the high pressure injection (HPI) design and associated flow rates. We require that you provide the following information:

a. Table of expected HPI flow (1 and 2 HPI pumps) in each of the four legs versus RCS pressure (2500 to atmospheric) considering the new cavitating 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 to the HPI flow rates assumed in the B&W LOCA analysis, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plants," May 7,1979, Volume 1. Figure 6.2.59.

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

(1) description of temporary flow indications and where they will ae 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 vill be taken. (4) range of installed flow instrumentation. (5) acceptance criteria f or flow rates at higher pressures. We require that this test confirm that the IMI HPI design provides adequate flow as assumed in Figure 6.2.59 of the B&W analysis (above). Provide your commitment to conduct a teet and submit the test procedure which will accomplish this purpose.

RESPONSE

Report Number GED 0005, High Pressure Injection Cross Connect, contains the information requested on the THI-l high pressure injection (HPI) design. Tables 2 and 3 of the subject report provides the expected HPI flows for one pump operations. Table 4 provides the expected performance for two pump operations. The analytical / empirical basis for these flow rates is contained in Sections 3.2 and 3.3 of Report GED 0005. These Tables and Sections of the report demonstrate that the HPI design providcs adequate flow as assumed in Figurc 6.2.59 of the B&W LOCA analysis, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plants". 1704 275 Am. 10

SUPPLDIENT 1, pART 3. QUESTION 1 (Cont'd.) Calculations to quantitatively define what reduction in flow rate was caused by the inclusion of the cavitating venturi are not available. However, from the calculations that have been made some qualitative assessments can be made. For one pump operatiens, the cross-connect design with venturis appears to result in a slight improvement in HPI perf ormance. The venturis appear to take the place of the resistance provided by the partially opened high pressure injection valves while providing a more even distribution of flow to each of the four HP1 legs. For two pump operations, the venturis do restrict HPI flow slightly with the amount of this reduction increasing as RCS pressure decreases. For example, at an RCS pressure of 600 psig, the venturis will restrict flow to about 84% of that achievable without the venturis installed. However, this still results in almost 87% more flow than required by the B&W LOCA analysis. The test abstract for the testing to be performed on the HPI system prior to return of TMI-l to power operations is contained in Section 5.0 of Report GED 0005. The information concerning test description is contained in this section of the report except for the basis for the 550 gpm " upper limit" acceptance criteria. Section 3.4 of the subject report contains the basis for the 550 gpm upper limit. When available, the test procedure for the subject testing will be provided. Am. 10 1704 276

'.~ -;.p.qq

   ,7 ATTACHMENT TO SUPPLEMENT 1, PART 3, ObESTION 1
   /RJ      .

SerW5CfB soo 211. nev 1 SYSTEM DESIGN DESCFJPTION FOR HIGH PRESSURE INJECTION CROSS-CONNECT PREPARATION ___ ' w DATE /AI?/!79 (

                                                             /     /

CONCURRENCE DATE APPROVAL DATE A000 0027 1704 27.7_

HIGH PRESSURE INJECTION CROSS CONNECT THREE MILE ISLAND UNIT #1 REPORT NUMBER: GED0005

                                           ,pd n Prepared by: y , f . ~YA~

I Jf F. Fritfjn d Metropolitan Edison Company Three Mile Island Unit #1 Middletown, PA 1704 278

TABLE OF CONTE!TTS Section Page 1.0 Introduction 1 2.0 Summary 2 3.0 Discussion 3 3.1 Background 3 3.2 Calculational Techniques 3 3.3 System Performance Under Cold Leg Break Conditions 3 3.4 Core Flood Line Break 4 3.5 High Pressure Injection Line Break 4 3.6 Normal Plant Operations 5 3.7 Transient System Operations 5 3.8 Core Cooling Using Only High Pressure Injection 6 4.0 Operator Action 7 S'. 0 Post Hodification Testing 8 5.1 Required Test Equipment 8 5.2 Procedure Abstract 8 5.3 Acceptance Criteria 9 6.0 References 10 1704 279

LIST OF FICURES . Figure No. Title 1 Flow Diagram for Pump "A" 2 Flow Diagram for Pump "C" 3 Simplified System Schematic of HPI Cross Connect LIST OF TABLES Table No. Title 1 High Pressure Injection Flow Requirements 2 High Pressure Injection System Performance With Valves MU-V16 A & B Open and MU-PlA Operating 3 High Pressure injection System Performance With Valves MU-16 C & D Open and MU-PIC Operating d 4 High Pressure Injection System Performance With Two Pumps and All Valves Open 5 Hibh Pressure Injection System Performance Under High Pressure Injection Line Break Conditions 1704 280

1.0 Introduction To improve the ability of TMI-1 to withstand the consequences of a small break Loss-of-Coolant Accident (LOCA), a change in the design of the high pressure injection system was developed. The subject change involves cross' connecting the "A" and "C" high pressure injection (HPI) legs and the "B" and "D" high pressure injection legs. The design is shown on CAI flow diagram C-302-661 Rev. IA-3 and detailed on CAI drawing E-304-666 Rev. IA-3. This report describes and presents the results of the flow calculations performed to verify the adequacy of the cross connec. design. As a result this report also represents the system description for the TMI-1 high pressure injection system. M 9 1704 281

2.0 Summary The criterion established by B&W for the small break analysis requires that 70% of the total flow for one HPI pump be injected into the unbroken legs _of the reactor coolant system. This criteria applies to a 2772 Mw thermal 177 fuel-assembly plant. For TMI-l with a licensed core power of 2535 Mwt, reference (a) indicates the 70% - 30% criterion can be relaxed in direct proportion to the amount that licensed power is less than 2772 MWt. Therefore for TMI-1, the acceptable flow split can be relaxed to 64% - 36%. The above criteria are applicable to all breaks except for those small breaks which occur as a result of a break in the high pressure injection line between the RCS and the first isolation check valve in the injection leg. The analysis presented in this report demonstrate that the proposed cross connect at THI-l can meet the B&W ECCS acceptance criteria. The analysis of this report further demonstrate that if the small break LOC /. results from a break in the HPI line, suf ficient high pressure injection flow will occur through the unbroken HPI lines to satisfy small break ECCS criteria. 1704 282

3.0 Discussion . 3.1 Background A mechanical piping cross connect has been accepted as a viable long-tern solution for additional corrective action on the small break LOCA. The basic. concept is derived from the BSAR 205 plants. The issue and scenario to arrive at the solution are extensive, well documented by references (d) thru (p), and will not be repeated in this report. The proposed design changes to be implemented at THI-1 are detailed by GAI drawings C-302-661 Rev. IA-3 and E-304-666 Rev. IA-3. The following discussion presents the system flow calculations performed in support of the subject design changes. 3.2 Calculational Techniques A simplified sketch of the proposed TMI-1 cross connect is shown in Figure 1 and Figure 2. Based on these layouts, a task was initiated to determine the performance of the proposed design under small break LOCA conditions. Included in this consideration were those small break LOCA's which could result from a HPI line break. . CAI's "PIPF" computer code was used to model the system. The PIPF Code is described in Topical Report CAI-TR-105NP-A. The subject report has been submitted to the NRC and accepted. 3.3 System Performance Under Cold Leg Break Conditions The initial performance of the TMI-1 high pressure injection system was evaluated with all RC loops assumed to be at a pressure of 600 psig. For this analysis the design did not include installation of cavitating venturis. The worst case flow slit occurs with only the "C" make-up pump running and under these conditions a 68% - 32% flow split was predicted to occur. B&W was asked to evaluate these conditionns and to investigate the possible relaxation of the 70% - 30% criterion based on the lower power level for TMI-1. In reference (a), B&W documented their conclusion that the 68% - 32% flow split' was acceptable. As further assurance of the acceptability of the TMI-] design, B&W was requested to provide the HPI flow assumptions utilized in their small break analysis code. Reference (b) and (o) transmitted this information. The criteria of reference (b) and (o) are contained in Table 1 and are ' based on undegraded pump performance. For the Small Break ECCS cal-culations, a factor of 0.9 was applied to subject flows to account for pump degradation due to wear. The resulting degraded pump performance is also presented in Table 1 and forms the basis

• for Figure 6.2.59 of reference (c). B&W was also asked to specify the maximum pressure difference which would exist between the unbroken cold legs and the cold leg containing the break. It was reported this value would be less than 4 psig.

Based on the above, CAI has performed calculations for each of the RC pressures specified in reference (b). For conservatism, the shortest leg was assumed broken and at a pressure 4 psig lower than the other legs. To account for the worst case single failure, only one high pressure injection pump was assumed operating. Undegraded pump performance was

                                                                                                     .      1704 283

assumed. For reasons discussed in Sections 3.5 cavitating venturi were incorporated into the original cross-connect design. The sizing of the cavitating venturi was based on pump *un-out considera-tions. Run-out flow for the TMI-l high pressure injection pumps is cons.idered to be slightly greater than 550 gpm since higher flows have not yet been demonstrated by test. The venturi were, therefore, sized to limit flow to 137.5 gpm (i.e. one-fourth of 550 gpm) when only a single pump is operating. Specifically, they limit flow.to 137.5 gpm when the venturi inlet pres-2re is 813.6 psia. At thic flow the vendor indicates that the non-recoverable pressure losses are 15% of the inlet pressure. The GAI calculations, therefore, assumed that the maximum achievable flow varied as the square root of the absolute inlet pressure and that the non-recoverable pressure losses varied as the square of the flow. Table 2 and 3 presents the calculated results of the TMI-l high pressure injection system performance under the above assumptions. Table 4 pre-sents the system performance when both pumps are running. The acceptance criteria was obtained by multiplying the values of reference (b) by 0.7 to account for the 70% - 30% flow split criteria of a 2772 FN plant and by the power ratio of TMI-l to the generic plant design (i.e. 2535/2772). As indicated by Tables 2 and 3, the TMI-l cross-connect design with cavi-tating venturi not only meets the above acceptance criteria but also meets the 7C/30 flow split criteria for a 2772 Mwt plant. 3.4 Core Flood Line Break The core flood line break establishes the maximum size acceptable for a cavitating venturi, Under these accident conditions, low RCS pressure occurs and high pressure injection is required. The cavitating venturi should, therefore, limit flow to less than run-out flow of a single pump. Run-out flow for the THI-l high pressure injection pumps is considered to be slightly greater 550 gpm since higher flows have not yet been demonstrated by test. As indicated in Section 3.4, the venturi has been sized based on limiting flow to 137.5 gpm when the inlet pressure is 813.6 psia. Based on this size venturi, calculations performed by GAI indicate that below 600 psig RC system pressure, the venturi will be in cavitation and flow will be limited to 551.5 gpm assuming the highest head pump is in operation. 3.5 High Pressure Injection Line Brerk One of the small breaks considered in this evaluation was that associated with the break of a high pressure injection line break. S,uch a break if it were to occur between the RCS cold leg and the first isolation check valve would res' ult in both a small break LOCA and a HPI line break. Calculations performed on the original design (i.e. the design without cavitating venturi) indicated that the HPI line break required the operator to isolate the leg containing the high flow. Such action goes against the operators normal tendancy and therefore presents the opportunity for operator error. As a result, the system design was modified as discussed above by the addition of cavitating venturi in each of the HPI legs. These devices will climinate the need for the operator to take action which goes against his normal judgement. 1704 284

Based on the venturi parameters presented in Section 3.3, the performance of the HPI system under high pressure line break conditions were analyzed. The results of the analysis are presented in Table 5. These results were, then, transmitted to B&W. Their review and conclusions are docu-mented in reference (o) and indicate that the design performance of the system will be acceptable. Table 5 summarizes the acceptance criteria contained in reference (o) for this break and demonstrates the acceptable performance predicted for the system. 3.6 Normal Plant Operations Due to space limitations at TMI-1, the hir,h pressure injection line cross connect can only be accomplished inside of the reactor building. The cross connect will, therefore, be installed downstream of the location where under normal operations reactor coolant system make-up was being supplied. This resulted in an unacceptable system design since due to the cross connect, oscillating make-up flow between the "B" and "D" legs could occur and result in high cycle thermal fatigue failure of the injection nozzles. To correct this problem the normal make-up injection point to the "B" HPI line was relocated as shown in Figure 3. Specifically, the line is to be extended into the reactor building through spare penetration 323 and connected into the "B" HPI line downstream of the cross connect. A check valve will be added to the "B" HPI line to prevent back flow of normal make-up water to the cross connect. The existing containment isolation valve, MU-V18, will precide outside containment isolation for the line. Inside containment isolation will be accomplished with a check valve. With the above changes, the normal system operations of the TMI-l make-up and purification system will remain unchanged. 3.7 Transient System Operations During transients which result in overcooling of the reactor coolant system, presaurizer level decreases and operator action is taken. This action currently consists of opening MU-V16B and starting a second make-up pump as necessary to restore pressurizer level. This action is required because normal make-up control valve, MU-V17, significantly restricts the maximum deliverable flow. The above operator action provides high flow rates without any thermal shocking of the HPI nozzles. With the installation of the cross connects, continued operator action in this manner would result in additional thermal shocking of the "D" HPI nozzle. Operations in this matter is not recommended since the stress calculations for the HPI line nozzle allow only 80 cycles of cold water injection to a hot nozzle (a nozzle without continuous flow). As a result, an alternate means for initiating high make-up flow to the "B" HPI leg has been provided by the design. Specifically, bypass valve, MU-V217, is to be installed around MU-V17. Valv.e, MU-V217, which is a high pressure injection valve similar to the MU-V16 valves, will be a normally closed, motor operated valve capable of being opened by the operator from the control room. As such it will provide the same function as the current MU-V16B and allow the operator to deliver a maximum of 450 gpm of make-up at 1800 psig RC pressure backpressure. Flow will only be through the "B" HPI nozzle and therefore thermal shocking will not occur. 1704 285

To provide indication of flow, a flow meter is being installed on the bypass line. The flow meter is a strap-on sonic flow meter manufactured - by Controltron Corporation. This flow meter will have a range of 0 to 500 gpm. The output of the flow devices will be transmitted to the control room where a meter will be installed to read flow directly. 3.8 Core Cooling Using Only High Pressure Injections The high pressure injection system provides a back-up means of core cooling in the highly unlikely situation where all secondary system cooling, including auxiliary feedwater is lost. To provide this back-up cooling capability, B&W analysis indicates that one HPI pump capable of injecting 216 gpm at an RCS pressure of 2500 psig is required. This assumes decay heat following reactor trip are at levels given by ANS 5.1 with a safety factor of 1.0. The TMI-l cross-connect design has been analyzed to ensure that the cavitating venturi due not restrict flow below 216 gpm. The results of the GAI analysis indicate that the system will be capable of injecting a minimum of 253.9 gpm at an RCS pressure of 2500 psig. Therefore the back-up method of core cooling using the HPI system will be available.

• 1704 286

4.0. Operator Action . Following actuation of High Pressure Injection, no specific operator action should be necessary to achieve proper flow and flow split condi-tions. The on,1y immediate action required by the operator is to verify that at least the minimum level of actuation has been achieved. This can be ac'complished using flow transmitters MU23-DPT 1 thru 4. The criteria for successful actuation is as follows:

1. Total flow must be greater than or equal to that shown in column B of Table 1.

2. If both HPI pumps are operating, two or more HPI valves must have full open indication except t'.ta t an indication that only valves MU-V16A and 16C or MU-V16B and 16D a re open represents an unacceptable two valve combination. The installed flow indicators should be used to confirm the valves have indeed open.

3. If only one pump is actuated. both MU-V16 valves in the train with the operating HPI pump must be full open. The installed flow indi-cation should be used to confirm the valves have indeed opened.

If the above criteria are met then the operator is assured of compliance with the ECCS acceptance criteria. If the above criteria is not met then multiple failures or other reasons have prevented proper automatic actu-ation. In this event the operator must diagnosis the problem and take corrective action (i.e. start idle HPI pumps, open closed MU-V16 valves, etc.). s 1704 287

5 5.0 Post Modification Testing The following tests of the high pressure cross-connect design are recommended prior to the return of TMI-l to power cperations. 5.1 Required Test Equipment

1. Wide range RC Pressure - RC3A-PT3, RC3A-PT4 and RC3B-PT3; range O to 2500 psig.

2. HPI Flow - MU23 - DPT1, 2, 3 & 4; range 0 to 500 gpm.

3. Temporary HPI Injection Leg Flow Meters - FX - 1, 2, 3, 4 (See Figure 3) consisting of:

a. Flow Display Computer: Controltron P/N 24IN - 2.5SS.375; Range 0 to 330 gpm.

b. Multiplexer - Manual Selection 4 channels Controltron P/N 242-10.

c. Transducer (4): Controltron P/N 240N - 2.5SS.375.

d. Cables (4): 25 ft each Controltron P/N 242 25.

4. Make-up Pump Discharge Pressure: MU22 PII, MU22 FI2, MU22 PI3; range O to 5000 psig.

5. Make-up Pump Suction Pressure: PX-412, PX-413, PX-414 5.2 Procedure Abstract d

1. Establish RC pressure at less than 600 psig and take adequate precautions for 3verpressure protection.

2. Ensure minimum recire. line is open on pump to be operated.

3. Start MU-PI A or MU-PIB and slowly open MU-V16A and MU-V16B. Maintain balanced flow.

4. Shut MU-V36 or MU-V37 to secure pump recire. prior to reaching 400 gpm total flow. Open MU-V16A/B to full open position.

NOTE The cavitating venturis were designed to prevent inadvertent pump run-out and cavitation without the need for throttling MU-V16 A thru D. Maximum expected flow below 600 psig RC pressure is 552 gpm. If necessary set MU-V16A/B to limit flow to less than 550 gpm. Reduce RC pressure to verify venturies are in cavitation as indicated by flow remaining constant. CAUTION Extended operations of HPI pumps under cavitating conditions should be avoided. Terminate test as soon as cavitation is observed or unstable conditions are observed. Do not allow pump discharge head to drop below 450 ft. Record flow at which unstable conditions are reached. , , 1704 7.88

CAUTION Do not operate any other make-up pump while minimum flow tecire. line is secured. Do not close MU-V16A and B while recire. line is isolated and pump is running.

5. Record flow through MU23-DPTl & 2 and through each sonic flow indicator.

Record pump suction and discharge pressure.

6. Reduce flow to less than 400 gpm ahd open MU-V36 and MU-V37.

7. Start MU-PIC. Open MU-V16C and MU-V16D and close MU-V16A and MU-V16B as necessary to maintain flow in each of the four legs.

Stop MU-PIC.

8. Repeat steps 3 and 4 for valves MU-V16C and D and MU-PIC. If necessary use MU-V64C for recire. isolation in lieu of MU-V36 and MU-V37.

9. Record flow through MU23-DPT3 & 4 and through each of the sonic flow meters. Record pump suction and discharge pressures.

10. Open or verify open MU-V36, MU-V37, and MU-V64C.

11. Secure testing unit 1 RCS pressure is raised to 1200, 1500, 1600 or 1800 psig. Testing at 1200 psig is preferred, however, testing at one of the other pressures is acceptable.

12. At 1200, 1500, 1600, or 1800 RCS pressure, start MU-P1A and or MU-PlB if not already started and open MU-V16A and MU-V16B.

s

13. Shut MU-V36 or MU-V37. Observe caution of step 4.

14. Record flow throufh MU23-DPT 1 & 2 and through each sonic flow indi-cator. Record pump suction and discharge pressure.

15. Open or verify open MU-V36 and MU-V37. Close MU-V16A and MU-V16B.

5.3 Acceptance Criteria Step 5 - Flow in any leg shall not exceed 30% of total flow. Total flow of all four legs shall be greater than 500 gpm and less than 550 gpm. Step o - Flow in any leg shall not exceed 30% of total flow. Total flow of all four legs shall be greater than 500 gpm a.nd less than 550 gpm. Step 14 - Flow in any leg shall not exceed 30% of total flow. Total flow of all four legs shall be greater than that specified in Table 1 Column B. In addition, if total flow of all four legs is less than 95% of the value specified in Table 2, an engi-neering evaluation of the data shall be requested. 1704 289

6.0 References (a) B&W (G. T. Fairburn) letter TMI-79-17 dated February 5,1979 letter to R. M. Klingeman, RE: Three Mile Island Nuclear Station - Unit 1 Small Break Analysis. (b)' B&W (G. T. Fairburn) letter TMI-79-74 dated May 21, 1979 to J. F. Fritzen, RE: HPI System Flow. (c) B&W Report " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant", dated May 7, 1979, Volume I. (d) Met-Ed letter GQL #0714 dated April 17, 1978 from J. G. Herbein to R. W. Reed / S. A. Varga (NRC). (e) GPUSC memorandum NF-279 dated May 2, 1978 from G. R. Bond to W. E. Potts. (f) B&W 1etter dated May 1,1978 from J. T. Janis to W. E. Potts, Re: THI-l Small Break Analysis. (g) Met-Ed letter GQL #0854 dated May 5, 1978 from J. G. Herbein to S. A. Varga (NRC). (h) Met-Ed letter GQL #0907 dated May 11, 1978 from J. G. Herbein to S. A. Varga (NRC). (i) B&W 1etter dated July 18, 1978 from James H. Taylor to S. A. Varga (NRC). (j) Met-Ed letter GQL #1254 dated July 24, 1978 from J. G. Herbein to R. W. Reid/ S. A. Varga (NRC). (k) Met-Ed letter GQL #1619 dated November 21, 1978 from J. G. Herbein to R. W. Reid (NRC). (1) Meeting Report dated December 12, 1978, Re: Licensee's Revised Proposed Modification to Eliminate Reliance on Prompt Operator Action Following A Small Break LOCA. (m) Met-Ed letter GQL #2031 dated December 21, 1978 from J. G. Herbein to R. W. Reid (NRC). (n) Met-Ed letter GQL #2072 dated December 29, 1978 from J. C. Herbein to R. W. Reid (NRC). (o) B&W (G. T. Fairburn) letter TMI-79-208 - dated December 21, 1979 to D. G. Slear, RE: Accessement of HPI System to Mitigate Small LOCA's. (p) Met-Ed/GPUSC Report In Response to NRC Staf f, " Recommended Require-ments for Restart of Three Mile Island Nuclear Station Unit 1", dated September 7, 1979 as supplemented and amended. 1704 290

                                                                                                             ~

FIGURE 1 ' Flow Diagram for Pump "A" , Leg "A" Leg "C" Le "B" Le "D" , , [ b ' L21=301.3 ft. L22=237.h5 ft. L23=196.19'ft. L2h=120.87 ft. -

                   / :(

FE-3Bk FE-386

                                            )(      FE-385 )(                FE-387      )                                  ,'

Lll*=159 98 ft. eq. . MU-V16A MU-v16B L12*=lh8.Tl ft eq. d = 2.125 in. d= 2.125 in.

                                                                                                                ~
                              )                                      MU23 )
     '                        I {MU23FE1                             FE2 I   (

C U

                                                                    >    RCP 3eal Injection
                          - L1=93.6 ft. eq.         -
                                                                                                                                         ~

d1=3.62h in.

                                   ~

MU-VThA L1A=121 98 ft. eq. d1A=2.62h in.

• MLL-VT3A 'I MU-PlA ,.

 *cxclusive of valve equivalent length

FIGURE 2 ' ' . ". 9.'.n

                                                                                                                        '..;~,' : ' .*l' .

Flow Diagram for Pu=p .

                                                          "C"                   .
                                                                                                                                  .        c'.. . : -
                                                                                                                                        ~'

Le6 A" Leg "C" "B" La "D" "

                                                                                                                       ~

Le A  ?* L21=353 75 ft: L22=185 ft.* L23=2h9.69 ft. L2'h=67.37 ft. h '."'.

                                                                                                                                              .,h FE-384) (            FE-386                                                                                 -                       -

FE-385} ( FE-387)! ( _ LH"= 6 ft. eq. ,. MU-V1dC' IG-V16D L12*=2 1 9 t. eq. MU23 ' E3 MU23) Eh J (

                                           .- - r    %

N # L1=86.52 ft. eq. , d1=3.624 in. MU-V7hC . LIA=121.98 ft. eq, .. dlA=2.62h in. . .. .- 13U-V73C _,'dM-s

                                                                                                                                         ' Y.-
                                                                                                                                      .,.1.....

MU-PIC. .. .

• exclu:ive of valve equivalent length .

1704 292 - i-

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Table 1 liigh Pressure Injection Flow Requirements Column A Column B RC Pressure, psig Pump Flow Pump Flow Undegraded, gpm Degraded, gpm 0 550 495.0 600 500 450.0 1200 437.1 393.4 1500 404.3 363.9 1600 391.4 352.3 1800 364.3 327.9 2400 260 234.0 2500 216.0 104.4 (1) Only for small break conditions other than a HPI line break. See reference (o) for flow assumed under HPI line break conditions. 1704 294

d w eo t l 92 8 4 9 9 8 4 5 aF l 19 5 6 8 4 4 2 7 ul 43 3 7 3 2 9 8 5 c a 55 5 4 4 4 3 2 2 laopt m CT g m) p1. g( d 1 1 8 8 6 2 4 3

           , e)                .          .

s r% 2 0 9 8 0 3 6 8 gi4 5 2 7 5 5 3 6 3 e u6 3 3 2 2 2 2 1 1 L q( e ) n R l g e a n k t i o o t r T a b d r n e g 64 8 1 0 9 2 9 7 e p U t me . . auL 1 9 6 9 2 1 0 8 0 O n l m 88 8 3 1 0 8 9 8 i uie 33 3 3 3 3 2 1 1 A cne 1 w l i r P o af h l

    -   l    C         T U    F             (

M d n n a o l i 30 3 3 1 7 7 2 5 n at m . . . . e ecp 41 8 6 5 4 3 0 9 p S eg 32 1 1 1 1 1 1 O j n B I A

                  "D       08         7         0      8      3      9    4     3 6               "

2 1 68 0 1 1 8 0 3 7 V g 22 3 2 1 0 0 7 6 E - e 1 1 1 1 1 1 1 L U L B M A m T s p "C 85 1 6 5 2 9 8 7 e g " . v 68 9 3 4 1 3 6 0 l n g 22 2 1 0 0 9 6 6 a i e 1 1 1 1 1 1 V L w h o t l "B 06 9 8 6 2 8 1 9 i F " . . . w 70 1 4 5 2 4 7 0 I g 23 3 1 0 0 9 6 6 e P e 1 1 1 1 1 1 c H L n a m "A 83 8 7 9 5 5 9 1 r " o 70 5 0 1 8 1 4 9 f g 23 2 1 0 9 9 6 5 r e 11 1 1 1 e L P m e "D t " s . y p 06 6 6 6 s 6 6 6 6 i S o 9 9 9 9 9 9 9 9 s g o 4 5 1 4 5 7 3 4 a I L 1 1 1 1 2 2 P i b H s r P "C o

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                   "                                                      f n          p i              0    0     0    0      0   0    0  0             0 o        0     0    0     0    0    0  0 e         o        6     2    5     6    8    4  5             2 r        L               1    1     1    1    2  2 u                                                               n s                                                               o s       "B                                                     i e       "                                                       t r                                                               c P          p   0    0     0    0     0    0    0  0              e o        0     0    0     0    0    0  0             S C          o        6     2    5     6    8    4  5 R         L               1    1     1    1    2  2              o t
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TABLE 4 IIPI System Performance With Two Pumps Operating and All valves Open RC Pressure in Psig HPI Flow in gpm Flow in Unbroken Legs, gpm Calculated Seal Calculated Required (1) Total Flow Loop "A" Loop "B" Loop "C" Loop "D" Leg "A" Leg "B" Leg "C" Leg "D" Injection Minimum (64%) gpm epm 0 0 0 0 199.1 199.0 199.1 199.0 51.8 597.1 352.1 84 8 600 600 600 600 199.4 199.7 199.4 199.7 42.3 598.5 320.1 840.5 1200 1200 1200 1200 201.2 201.0 200.9 201.7 29.8 603.1 279.8 834.6 1500 1500 1500 1500 188.4 196.3 197.4 204.9 23.2 582.1 258.8 810.2 1600 1600 1600 1600 181.9 189.3 190.5 200.3 22.5 561.7 250.7 784.5 1800 1800 1800 1800 169 175.9 177.1 186.2 20.9 522 233.2 729.1 2400 2400 2400 2400 119.7 124.7 125.6 132.1 15.2 370 166.4 517.3 2500 2500 2500 2500 108.6 113.2 114.0 120.1 13.9 335.8 138.2 469.8 N CD b N

-411) Refer to Section 2.0 for basis.

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                      "                                                       ort n                                                                    un i           p  0      0        0      0      0       0 sd   e o                                              0    i            m 0        0      0      0       0     0      s wu e          o         6       2      5      6       8      4      aoc r        L                   1      1      1       1      2 u                                                               bl o s                                                                  fd s        "B                                                      r e        "                                                       od e r                                                               f er P         p   0      0       0      0      0 ra C

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ATTACINENT TO SUPPLEMENT 1, PART 3, QUESTIONS 1, 2 6 3 Babcock &Wilcox ,,,,, o,,,,e,,,n o,co, P.O. Box 1260, Lyn:h:urg. Va. 2450s Telephone: (804) 3847 5111 December 21, 1979 TMI-79-208 Mr. D. G. Slear (2) TMI-1 Project Engineering Manager GPU Service Corporation , 260 Cherry Hill Road Parsippany, NJ 07054

Subject:

Assessment of HPI System to Mitigate Small LOCAs

Dear Mr. Slear:

We have completed our comparison of the Gilbert Associates predicted TMI Unit 1 HPI flows to results of previous HPI line break analyses for similiar 177 FA plants. Our comparison shows that the delivered flows demonstrate conformance with the criteria of 10CFR50.46 for small breaks. The results of our comparison are attached for your information and use. If you have any giestions, please advise. , Very truly yours,

b. ), d G. T. Fairburn.'

Service Manager GTF/cw - J. G. Herbein cc: L. L. Lawyer G. P. Miller R. S. Harbin

• J. F. Fritzen J. J. Colitz W. E. Potts bcc: g, g Ellison R. F. Wilson - GPUSC G. H. Olds R. W. Heward - GPUSC , b* R. Pletke F. R. Faist H. A. Bailey J. C. Lewis - Phila. Sales R. C. Jones M. G. Gharakhani Rec. Ctr. NSS-5/T1.2 1he Babcock & W.tcot Com: an. i Es abished 1867 1704 299 e

ASSESSME!!T OF MODIFIED TMI-l HPI SYSTEM TO MITIGATE SMALL LOCAs D D

                                                                                           *DWyn I. Introduction                                                                             d Modifications have been made to the T!!I-l HPI system to improve its capability of mitigating the consequences of small breaks in the primary system. These modifications include cross-connecting of the HPI system, which allows one HPI pump to feed all four cold leg pump discharge pipes, and cavitating venturis, which minimizes the loss of the HPI fluid in the event of an HPI line break.
    <         The capability of the revised HPI system has been reviewed to assure that the delivered flows are sufficient to insure compliance to 10 CFR 50.46 in the event of a small break.

• Reference 1 provides the flowrates delivered to the RCS by the Tf11-1 HPI system for a small break in the pump discharge piping and an HPI line break. These flows have been compared to the values assumed in the 177-FA lowered loop plant generic 2,3 The results of that review, provided belew, demonstrate that the TMI-l HPI system.provides more flow to the RCS than used in the analyses which confoms to 10 CFR 50.46. Thus, the TMI HPI system is acceptable.

II. Comoarison of TMI-1 HPI Flows to Arialysis Flows

             ~ A. A generic small break LOCA analysis has been performed for the 177FA lowered loop plants and is documented in reference 2. That analysis assumed a nominal core power lever of 2772 MLT and is conservative for the TMI-l plant which has a nominal power of 2535 MWT. In the analysis, the break was assumed to be located between the HPi nozzle and the reactor vessel inlet nozzle. Since a
        .            portion of the injected HPI would be lost directly out the break for breaks in the pump discharge piping, this break location yields the most severe
   , y:.  ,.         consequences and is chosen for analysis to demonstrate conformance to 10 CFR 50.46.

P 3:g' T .

      **;            In the analysis, only one HPI train was assumed to be initiated by the ESFAS.
                    , Initially, this results in only 505 of the HPI being delivered to the RCS for core cooling, and the other 500 is lost through the break. At 10 minutes after ESFAS, the operator was assumed to cross-connect the HPI lines, thus allowing 70% of the HPI to be delivered to the RCS for core cooling, and only 303 is lost to the cqntainment. The worst case for the above mentioned analyses was the 0.07 ft.' break, which resulted in a minor core uncovery and a peak cladding temperature of 10920F. This is well below the 22000F        ,

criteria of 10 CFR 50.46. The HPI flows assumed in the analysis to be delivered to reactor vessel for core cooling are listed in Table I along with the flows which will be delivered to the RCS by the T!!I-l plant HPI system for pump discharge breaks. As shown, the TMI-l delivered HPI flows for core cooling are in excess of the values assumed in the analysis, for all pressures and time periods. It should also be noted that no operator action is necessary for the IMI-l HPI system, since the HPI system modifications already nelude the cross-connection of the injection lines. Therefore, the HPI flows for the IMI-l plant are adequate for mitigating small break LOCAs within the criteria of 10 CFR 50.46 1704 $00

B. HPI Line Break DTD "D T%f

                                                                            #      d      -   -

Since the HPI system performar.ce is altered by a rupture of one of its lines, an llPI line break is considered separately from other small breaks. With the break in the liPI line, the mejority of the HPI flou is lost directly out the break and only a small portion of the llPI is available for core cooling. Thus, operator action by 20 minutes is generally required to isolate the breken line, and thereby increase the flow delivered to the P.CS. Cavitating venturi have been installed in the TMI-l HPI lines to restrict the flow through the broken HPI line and increase the amount of HPI fluid available for core cooling. Eccause of the increased flow available for core cooling, no

 <                  operator action is required for the THI-l HPI system.

An HPI line break has been analyzed for the 177 FA lowered loop plant operating at a ner.inal core power of 2772 MWT3 assuming operator action

                   .at 20 minutes to ~ isolate the broken HP1 line, the analysis shows that the core remains covered through the transient, and the cladding temperature is maintained within few degrees of the coolant temperature.

Thus, compliance to 10 CFR 50.46 is demonstrated. Table II lists the HPI flows assumed in the HPI line break analysis, along with flo is that the TMI-1 HPI system would deliver for this accident. As shown, prior to the operator's isolation of the broken line at 20 minutes the TMI-l flows are in excess of the ass 0med flows. After the operator isolates the broken line, the primary system pressure remains below 1100 psi for the remainder of the transient. For this pressure rance, the THI-l HPI system will deliver more flow than that assumed in the analysis. Therefore, the TMI-l HPI system will deliver more flow than that assumed in the analysis. Therefore, the TMI-l HPI system will mitigate the consequences of an HPI linebreak within the 10 CFR 50.46 criteria. III. Conclusion . Modifications have been made to the TMI-l HPI system to reduce the reliance on operator action for a small break. These modifications, which consist of cross-connecting of the HPI lines and installing cavitating venturis in each line, have been reviewed to determine if the delivered HPI flows are sufficient to control small breaks. By comparing the TMI-l delivered flows to the values "

   .         assumed in analysis performed to demonstrate conformance to 10 CFR 50.46, it was shown that the TMI-1 HPI system will provide sufficient flow to the RCS, in the event of a small break, to insure conformance with the criteria of 10 CFR 50.46.

• 1704 301

           ' '"       m=   m  ~

y

                                             ~~
                                                           ,w- -

TABLE I HPI FLOWS DELIVERED TO RV FOR PUMP DISCHARGE BREAKS I Pressure Assumed Assumed TMI-1 psig HPI Flows HPI Flows Flows The First 10 Min. After 10 Min, gpm 0 257.5 600 385 iI 398.5 225.0 350 398.5 1200 190 306 1500 ' 345.4 171 283 31 7.8 1600 164 274 1800 307.6 150 255 285.4 2400 130 182 ' 202.5 e>

                                                                                                      .3                                     -
                           ,                                           TABLE II HPI FLOUS DURING HPI Lit 1E BREAK I

Pressure Assumed Assumed psig THI-1 HPI Flow HPI Flow Flows The First 20 Min. After 20 Min. ODm CDA 003 0 350 368 E00 3FJ]T-~

     .                                              293.7                                      '33.5                       398.5 1200                         213.7                                      289.5                     -

303.4 1500 169 262.5 1600 251.1 150 253.76 1800 231.4 - 112.5 235 1 91 .1 1704 302  ! 9 _ _

                                     .W.               -
                                                                               . -    ,--"ws~       I.i # EN -- --- -         -     -

REFEREllCES s .

1. Letter R F. Ely and B. M. Rogers (GAI) to J. F. Fritzen (f'ET ED) ,

October 11, 1979

2. Letter J. H. Taylor (80.1) to S. A. Varga (flRC),

July 18,1978 I

3. Midland 1 and 2 FSAR, section 6.3.2.2.9.2.

e e e d e b e O O 9 e 9 e 1704 503 0 e e

               "~--p--           ,_ _ .

SUPPLEMENT 1, PART 3 QUESTION:

2. Your response to question 36a does not provide sufficient analytical justification for adequacy of the 64/36 flow split for an HPI line break or your statement that RCS pressure will not expend significant time above 1500 psig for a spectrum of HPI 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 HPI line breaks (between the RCS and the check valve nearest the RCS).

R2SPONSE Justification for the adequacy of the TdI-l HPI syste= design under an HPI line break is contained in Section 3.5 of Report GED 0005. B&W 1etter IMI-79-208 (i.e., reference (o) of the subject report) has also been included. 1704 304 Am. 10

SUPPLEMENT 1, PART 3 QUESTION:

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 HPI line break (limit flow out break such that sufficient HPI reaches core). It is our position that the brief test description does not adequately cover the conditions which would result from an HPI line break.

Also, provide detailed drawings, data, and specifications for the cavitating venturis.

RESPONSE

Information regarding the .avitating venturi design is contained in Section 3.3 of Report GED 005. The predicted performance of these venturis under HPI line break conditions is contained in Section 3.5 and Table 5 of the subject report. Copies of the purchase order for the venturi (P.O. # 86530) and the vendors drawing of the venturi are attached. Additional tests will be performed to demonstrate the venturi performance under both cavitating and non-cavitating conditions (See Section 5.2 of GED 0005 for test abstract). These tests will confirm the design performance of the venturi under all possible flow conditions and therefore, will adequately cover the conditions which could result from an HPI line break. 1704 305

V' .:. *

  7 I--        I         ,.         .                       SUPPLEMENT.1, PAPT 3. QUESTION 3                                                                          Page a ci 3
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ACCOUNTS PAYABLE DEPT.. P.O. BOX 542. RE ADING. P A. ~960: e DEllVERY REQUESTED q rl.~riEE'iliiEn ' v f!ILOSA: DELIVERY VI A 2/5/80 iSee Below .X i Best Method VENDOR NO. ST.R f.1 ,j. , C ,

                                                                                                 ! P 'ilE TXNT C] BLANKET ORDER NO. i i                                                                      .

l s I i l ITEM OUANTITY l Ull STOCK SY'/l/ni No DESCRIPTION l PRICE l CONFIRMATION - DO NOT DUPLICATE l All packages shipped against this order must reflect part name, part I number and Met'-Ed purchase order number. The vendor shall assure that the mater and long term ,ial isstorage. packaged to prevent damage during shipp1ng, receiving 1 4 Ea. Cavitating Venturi - with Butt veld ends for 2 " Schedule. 160 pipe material to be low carbon (0.04% max.) *austenitic stainless steel (316L or 304L) certified to an ASME or I ASTM specification. Minimum wall thickness shall be that specified for 2 " cchedule 160 pipe. Venturi will flow 137.5 GPM, 40 F Water, at 813.6 PSIA upstream pressure with downstream pipe broken, and will flow 115.4 GPM at 1310 PSIA upstream and 122.4 PSIA downstream, a For any preposed material with greater than 0.04% maximum carbon, vendor shall submit the Certified Material test report for OWNER approval prior to fabrication. Om y .- . QA REQUIREMENTS: oo oj ,J_k ,

1. The Vendor shall identify each item by showing part name, part number, item number, and Met-Ed purchase order number.

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2. READ' G.PA.

T he attach. o Acknowledomeni must t e seom ce M. reinened swomotiv. '

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2. i ne attached Ack noateegment must be s gned & etu.nett p.omptiv. ) RE ADING. .

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i {. M AIL INVOlcES IN TRIPLICATE TOGETHER WITH Wav2fL'. To ACCOUNTS PAY ABLE DEPT.. P.O. BCx 542, RE AolNG. PA. 1960; DELIVERY REQUESTFO .? r in n linrn v ~ N ' '.O S C DELIVERY VI A 2/5/G0 . . _ _ See Below. .uX. Best_Eethod VENDOR NO i T' ,f. ,tL g ix14 C b'.M MET ORDER MO.

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3  ! I t  ! I i l . I I l- e ITEM OUANTITY l U/l STOCK SYVHOL NO DESCRIPTION PRICE

6. The provisions of 10 CFR 21 apply hereto. The Con-tractor shall immediately inform the Company's Vice President of Generation in writing if it obtains in-formation reasonably irdicating that the Plant or service or materials or a basic component delivered to i the Owner for the Plant (a) fails to comply with the l l

Atomic Energy Act of 1954, as amended, or any applicable j rule, segulation, order or license of the Nuclear l

@                                                          Regulatory Commission ("NRC") relating to substantial ;

safety hazards, or (b) contains a defect, wnich could i create a substantial safety hazard, unless it has actual knowledge that the NRC has been adequate"; i i informed of such defect or failure to comply, 4.1 as

                                    '                      required by Part 21 of 10 Code of Federal Regtlations l           ("10 CFR 21"), and shall simultaneously furnish to the Company's Vice Preside t of Generation copies of any notification given by y.ne Contract'or to tte NRC pursuant to 10 CFR 21.

7. Vendor shall submit one reproducible drawing for OWNER approval prior so fabrication.

                                                   " Seller / Contractor / Vendor shall furnish all apprcved Quality Assurance and Quality Control procedures and docu i mentation stipulated in the Specification / Purchase Order /

Contract prior to shipment. One (1) set of all required documentation shall be included with each shipment. ~ Failure of supplier to comply, shall result in vlthholding g'j]@ payments until all requirements are fulfilled." IIEIER ALLINQU ES T . ,y

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1. Oe dee No.. F .0 B. Point end Pr es e to .msicae on mII s aevnec es mois A c k no *ieoe men t s. RE ADIN 'A.

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Lancaster, Pa. PPD i. ADD 86530 Sc52L t/3,0 ,

                                                                                                                                                                .' yq.m.7.e y      :.10/1.0/79   .. .                     u..

ISSU L D SHIP TO METROPOLITAN EDlSON COMP / I i THE PEPJIUTIT CO. (cent'd) SEE PAGE 1 I M All INVOICES IN TRIPLICATE TOGET**ER .elT es W AYSILL i l ACCOUNTS PAY ABLE DEPT., P.O. BOX 542. RE ADING. P A. 196* DELIV8iRy rtEQUES'ED n. r r,.t'l V I) *!N LO".A- DEllVERY VI A 2/5/80 See Below .X I Best Method VENDOR NO. _ _ j , " t >rt' t

                                                                                      ' CC        TP INN ! TdiI RLANKET ORDER t;O.
                                                                !                     i      i           l               1                       i u - . ..   .- . . _          + - - . .

ITEM OUANTITY I Uli STOCK sYm til N') DESCFitPTION ' PRICE i PAY;IE::T: e i The Permutit Company shall be reimbursed by the Purchaser ', l for the performance of all requirements associated with i i this Purchase Order on the basis of a firm fixed lump sum i

                                           !             price . . . . . . . S25,060.00/                 Lot tiet. Subject to the                                                    l l             following conditions:                                                                                                      ',

I I

      )                                                  A.) Permutit will supply four (4) Cavitating Venturies                                                                       ',
     /                                                         no later than 2/5/80.                                                                                                ;

B.) Delivery Prior to 2/5/80 - Permutit will be payed an e additional $100.00 per week per Venturi delivered i l before 2/5/80. i ' C.) Delivery after 2/5/80 - Permutit will deduct $100.00 i per week per Venturi for not meetin'g the delivery ' I date of 2/5/80. l D.) Penalty / Bonus will be calculated only on a weekly i basis. Penalty / Bonus will not be pro-rated for days ! l early or late, j l

                                                       ' E. ) All drawing authorizations are to be approved prior                                                                       {

to delivery of bulk materials. Permutic will be i 11abic for penalty conditions only if delays in l shipment is due to Permutic's failure to perform. I

                                                       , TErJ1S OF PAYMEt;T:

1704 309 i The Permutit Company shall be paid net 30 days from . - - _ - i . . . . . . . . REf.ER ALL,1NOUI > T :l g ,,,,

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                                                                                                                                                                  . .. . 4 lEI                        ,          d.)

( *) _ . . . . _ . . . ~ . . . PENNA. SELECTIVE SALES AND,USE TAX DIRECT PAY PERMIT NUMDER 00135 ~ IMPORT ANT NOTICE METitOPOt.pfd7 EDI COMPA

1. Orde. Peo . F .o U. Poen t anss re .. . 3a ppe.ee on .se t in o.. es an<e f A c k no.. *c<toasenis. RE . DING. '.

            ?. t e., ati hed Ac a nows.i+, nent n uni t.c s.gneis & ec tuene.t p nenettv.

3. Teoch neuveries w.it not be accetste.s m' tee J 30 8' M. or on Coe 5.*n y j/, Yj

) desirinatest mehdavs. unless othe w.w soci ef es t. Aas c enttitions appeseeng on eeveene vote .$er a pet of this ceder. Y

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c' ._ nge . .. ., k-L M .e .-3d * 'd' --l

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XXMX.XXXX 717-948-8000

                                                                                                                                                                          -<,.q 1;/30                Lacaster, Pa. PPD L ADD.                                          86530                    #;ig.a.%                        86330 10/10/.79p ,s :                         tr n                                     .,i-                                          tu.-                          .~'e,y*          ,            ,

13 S U !JI . i SHIP TO I f.1El ROPOLITAN EDISON CO..*PANY I i THE PERMUTIT CO. (cont'd) SEE PAGE 1

                         , _ _ _                                                                             MAIL INVOICES IN TRIPLt0 ATE TOGETHER WITH WAYBILL TO ACCOUNTS PAY ABLE DEPT.. P.O. BOX 542, RE ADING P A. 19603 D E ll_V._E R Y R E QU.E._ST E D
                                                                                           " r'    f:I-'.".'i n 8 f )

• f! LOSA' DEllVFRY Vf A .

2/5/80 , See Below X 3 Best Method

        ' VENDOR NO.                                . . _

_i _S T R '.', l,;iP ~7 ' f! N TXN C ~ BLANKET ORDE R NO. I i ITEM OUANTITY l Ull [ STOCK SYMBOL No. DESCRIPTION i PRICE I Purchaser's receipt of verified invoice. ,

       .                                    l               CORRESP0h'DENCE & TRANSMITTALS:                                                                                           ,

I l l l All correspondence and transmittals relative to this I I purchase order shall refer to the Purchase Order No. and i be addressed to the following: 9 [' Metropolitan Edison Company P.O. Box 480 l j Middletown, Penna. 17057 Attn: T. T. Reilly, Whse. # 2 l 1 Coordinate drawing approvals with Met-Ed's Mr. Dave l Huffman, (215) 929-3601, Ext. 6544 i l Confirming verbal arrangements made by T.T. Reilly (GPU) & your S. Andersen on 10/9/79. i

                                                          . Reference telex messages dated 10/4/79 and 10/10/79.
                                                                                                                                                                                      ,                              I
                                                                                           'O                                       n      '                 -
                                                       ,                                                     O           ,

9 I MS J l J Ia, : ' a I 1704 510  : 1EFER ALLINQUI T P:j, g ,,,...,.y.

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                                                                                                                                                           ,            ,  , p,;y i                 l     ,y.3 I                                                                           . . . . . . . . . _ .

TTR:tml / , JO ll [ l PENNA. SELECTIVE SALES AND USE tax DinECT PAY PERMIT NUMBER 00135 _f IMPORT ANT NOTICE METHOPOLIT ED N COMPA

    't.       O e de. No..

• O.D. Posnt and Pe .< , to ppene on .01 envoers ..n

                                    .                                                           1 A c k nonse.*g vient s.                                                                                                                             str OtNC P ,

2. T he armhe.1 Acknowleden ent must t,e signe.t F. eesueneJ pec'mptiv, i

3. Truck nehwer.e w.tr not ee accepted stie, 3 30 P M. r e on Como.inv i j ees.gnated holidays. waless othee wise soe. it.ed. 'e A

d. All ce=nditions_ aopea_ ring cp eeveese s.de_ ave a pa'l of thes order '

                                                    ,                                                                                      b ~ ~ " " " * " , *. , , ,             , , , , , , ,., , , -b' Oa*.- J. E. Kunkel

SUPPLEMENT 1, PART 3 QUESTION:

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

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 EFW system n'as brought into question the capability of. tha_EEW.J. low control valves to meet the single f ailure criterian 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 EFW flow control valves. (No credit can be taken for the manual control stations in your analysis.)

RESPONSE

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 EFW system. This upgrading 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. d Am. 10 1704 311}}