ML19305C706

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Flow Detectors for Pressurizer Code Safety & Pilot Operated Relief Valves.
ML19305C706
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 03/03/1980
From:
GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML19305C698 List:
References
RM-10, TDR-139, NUDOCS 8003310287
Download: ML19305C706 (32)
v • d • e


Text

ATTACHMENT 2 TO APPENDIX 2A ,

O b f3g7g}gg TCR NO. 139 REVISION NO.

PAGE 1 CF 10 TECHNICAL DATA REPORT PROJECT NO. _.TMI-l h TMI-l Restart DEPARTMENT / SECT'ON Electrical Power & Instru.

RM-10 RELEASE DATE REVl3'ON DATE DOCUMENT TITLE: Flow Detectors for Pressurizer Code Safety and Pilot Operated Relief Valves.

DATE I' APPROV;;L(S) SIGN ATURE. DATE ORIGINATOR StGNATURE / /

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  • DISTRIBUTION  !, ABSTRACT:

In order to detect flow through either of the pres-T. G. Broughto:t surizer cod 2 safety valves or the pressurizer PORV, flow G. R. Capodanno measurement devices will be installed downstream of each of the D. K. Croneberger valves. These consist of differential pressure transmitters R. W. Keaten connected across elbows in the valve discharge piping.

I. D. Porter D. G. Slear The conclusion of this report is that this scheme

/' C. W. Smyth will provide satisfactory signals for detecting flow through V E. M. Steudel the pressurizer safety and relief valves.

A. L. Van Riper l E. G. Wallace l l

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  • COVER PAGE ONLY 8003310g(}

TDR No. 139

~

1. PURPOSE AND

SUMMARY

NUREC-0578 ("TMI-2 Lessons Learned") requires that positive position indication be provided in the control room for the pressurizer code safety valves (RC-RVlA,B) and the pressu-i:cr pilot operated relief valve (RC-RV2). This has been accomplished by providing flow detectors downstream of each of the three valves. These detectors consist of differential pressure transmitters connected across elbows in the valve discharge piping. This is shown schematically on GPUSC drawing D-601-110-001. The purpose of the report is to determine whether the D. P. transmitters will produce detectable signals under the flow conditions of interest and to provide a basis for specifying set points and ranges for calibrating the instruments.

It should be emphasized that this design is intended to give the operator a reliable indication that a safety valve or relief valve is open. It is not intended to provide accurate quantitative data on flow rate. The D.P. signal for a given mass flow will be substantially dif ferent, depending upon conditions in the pressurizer and the Reactor Coolant Drain Tank. In addition, in order to provide sensitivity to relatively low flons, it will be necessary to allow the transmitter to go off-scale for high flows.

2. METHODS The centrifugal force of a fluid flowing through an elbow creates a difference in pressure between the inside and outside of the elbow. This may be used as a means of measuring flow through the elbow. The method used for calculating the ps expected dif ferential pressure for specific flow conditions is explained in

\,,). Appendix A (Calculation Sheet No. 110lX-322B-001). It is the method referenced in " Principles and Practice of Flow Engineering" by L. K. Spink.

The fluid conditions in the discharge piping which are required for these calcu-lations have been determined by Gilbert Associates and are tabulated in TDR-170.

3. EVALUATIO,N The critorion for an acceptable design is that the D.P. transmitters will furnish a detectable signal at 10% of full-open flow. The calculations show that this criterion can be met. The lowest signals of interest are for 10% flow with a bubble in the pressurizer at 2200 psig. At higher pressures or with the pres-surizer solid, the signals will be greater. It will be assumed that the Reactor Coolant Drain Tank is at 70 psia since this produces a lower signal.
4. RESULTS The complete results of the calculations are given in Appendix A. A tabulation for the flow conditions of interest is given in Table 1.

-5. CONCLUSIONS

( It is concluded that the system will operate satisfactorily if each of the three j D.P. transmitters and its associated bistable are calibrated as follows:

) Transmitter span: 0-400" water (4-20 m.a.)

Bistable setting: < 20" water (4.8 m.a.)

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1 Table 1

. Diff.

PZR RCDT R2ss Press.

PZR Press. Press. Flow Signal Valve Condition (PSIG) (PSTA) (ibm /hr) (inches H,0) -

RC-RVIA, B Sat. Steam 2500 14.7 322,370 1054 Full Flow 2500 70 322,370 1054 2200 14.7 94,750 894 2200 70 94,750 894

" " " " 32,347 93.5 10% Flow 2500 14.7 " "

" " " " 32,347 47.1 2500 70 2200 70 27,528 36.9

" " 654,260 Sat. Water 2500 14.7 1601 Full Flow

" " " " 654,260 2500 70 1601 2200 14.7 634,510 1449 2200 70 634,510 1449 65,426 10% Flow 2500 14.7 185 2500 70 65,426 106 2200 14.7 63,451 171 2200 70 63,451 92.5 RC-RV2 Sat. Steam 2500 14.7 110,960 868 Full Flow 2500 70 110,960 868 2200 14.7 94,750 752

. p " " " "

2200 70 94,750 700 d " " " "

2500 14.7 11,096 74.1 10% Flow

^

2500 70 11,096 30.2 " -"

2200 14.7 9,475 59.9 2200 70 9,475 22.8 225,200 Sat. Water 2500 14.7 1230 Full Flow 2500 70 225,200 1230 2200 14.7 218,400 1139 2200 70 218,400 1139 2500 14.7 22,520 134 10% Flow 2500 70 22,520 67.5 2200 14.7 21,840 125 2200 70 21,840 59.1 i

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i,j Static Pressure @ Tap (PSI.\) Static Quality e Tap Mass Flow Rate ji RCDT(1) Pressure @

Case Description 14.7 PSIA 70 PSTA 14.7 PSIA 70 PSIA (1bm/hr)

'f:.

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A Saturated Steam

p. . 1. Full Flow S
a. 2500 PSIG 261 261- 0.86 0.86 110,960

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b. 2200 PSIG 226 243 0.89 , 0.89 94,750

[." .1.0(344*F) 40,590

'.fij c. 1000 PSIC 104 124 1.0(331*F) .

300 PSIG 37 77 1.0(308*F) 1.0(343*F) 12,565 d d.

' .t  !);.l 2. 10% Flow 11,096

.L.'l a. 2500 PSIG 29 74 0.92 0.90 74 0.95 0.93 9,475 Ik b. 2200 PSIC

c. 1000 PSIG 27 20 70 1.0(294*F) 1.0(325'F) 4,059 N.' 70 1.0(322*F) 1.0(345'F) 1,257
i'? d. 300 PSIG 15 Tl '
3. 10% Valve Rated Flow 30 75 1.G(307'F) 1.0(343*F) 10,000

'M a. 300 PSIG il B Saturated Water f,f [

e

1. Full Flow .

0.40 225,200 l ,.c a. 2500 PSIG 361 361 0.40

331 0.36 0.36 218,400

., b. 2200 PSIG 331 157,650

,j c. 1000 PSIG 179 179 0.23 0.23 95 0.14 0.12 88,573 f.. d. 300 PSIG 74

'i. 2. 10% Flow 22,520

.f\ a. 2500 PSIG 39 76 0.53 0.50 0./49 0.46 21,840

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-: b. 2200'PSIC 1000 PSIG 36 24 75 72 0.35 0.30 15,765 2'

h;-

c.

d. 300 PSIC 18 70 0.22 0.14 8,857 f, 3. 10% Valve Rated Flow 22,072
a. 300 PSIG 24 72 0.20 0.14 i
o. h, .  !

C Cold Water .

'5 ' 1. Full Flow .

1,224,000 RCD' i 'i .,

a. 300 PSIG 139 205 0.0(200*F) 0.0(200*F) 1,148,400 RCD"

.j, (1) Reactor Coolant Drain Tank .

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Table 1 - 6" Line Static Pressure G Tap (PSIA) Static Quality G Tap RCDT(I) Pressure G RCDT Pressure 0 Hans Flow Rate case Description 14.7 PSIA 70 PSIA 14.7 PSIA 70 PSIA (Ibm /hr)

A Saturated Steam

1. Full Flow
a. 2500 PSIG 356 356 0.85 0.85 322,370
b. 2200 PSIG 308 308 0.89 0.89 275,275
c. 1000 PSIG 143 148 'O.99 0.99 117,925
d. 300 PSIC 45 80 1.0(309'F) 1.0(341*F) 36,494
2. 10% Flow
a. 2500 PSIG 38 78 0.91 0.90 32,237
b. 2200 PSIG 33 75 0.94 0.93 27,528
c. 1000 PSIC 23 72 1.0(293*F) 1.0(325*F) 11,793
d. 300 PSIC 18 70 1.0(323*F) 1.0(344*F) 3,649
3. 10% Valve Rated Flow
a. 300 PSIG 40 78 1.0(309'F) 1.0(343*F) 31,170 B Saturated Uater
1. Full Flow
a. 2500 PSIG 433 433 0.38 0.38 654,260
h. 2200 PSIG 403 403 0.34 0.34 634,510
c. 1000 PSIG 244 244 0.20 0.20 458,000
d. 300 PSIG 86 105 0.13 0.11 257,255'
2. 10% Flow
a. 2500 PSIG 46 80 0.52 0.50 65,426
b. 2200 PSIC- 44 79 0.49 0.46 63,451

'c . 1000 PSIG 28 74 0.35 0.30 45,800

d. 300 PSIC 19 70 0.21 0.14 25,726
3. 10% Valve Rated Flow
a. 300 PSIC 29 73 0.19 0.14 65,015 C Cold Water
1. Full Flow
a. 500 PSIG 139 181 0.0(200*F) 0.0(200*F) 2,.;8,400 RCDTG14.

2,602,800 1:CDTG70.4 (1) Reactor Coolant Drain Tank

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cwP1rRoig TAELE 3.1-1 (Cont'd.), pig 3 2 Of 2 PROCEDURE NO. TITLE, PROCFIAIRE No. TITLE 1300-31 NSRWP and Valve Fur.ctional Test op 1105-1 helear Instrumentation EP 1202-35 Loss of Decay Heat Removal OP 1105-2 y Reactor Protection System EP 1202-14 Loss of Flow OP 1105-3 I

iSafeguards Actuation System SP 1300-3J NSCCWP and Valve Functional Test OP 1105-4 Integrated Control System EP 1202-29 Pressurizer System Failure OP 1105-5 Incore Monitoring System OP 1103-11 Drain and N2 Blanketing RC System or 1105-6 RCS Non-Nuclear Instruraentation SP 1300-3K RB Emergency Clg. Functional Test SP 1303-5.4 E W Pumps - Cancelled by PCR 79-493 EP 1202-36A Loss of Inst rument Air - B/U Air Avail. 1106-1 Turbine Generator EP 1202-365 Loss of Instrument Air - No 5/U Air Avail. 1303-11.22 M.S. Isolation Valves SP 1300-3L Screen Wash Pump Functional Test 1107-1 Normal Electric OP 1103-2 Fill and Vent 1106-2 Condensate System SP 1300-3M Screen House Vent Pump Funct. Test S101 F.W. Pump Turbines EP 12 2-2 Sta Blackout 1106-3 Feedwater System EP 1202-2A Sta Blackout with loss of Both Emer. Diesels S105 Exercise W Pump Emergency Covernor SP 1300-3N Chilled Water Pump Func. Test _1107-2,__ Emergency Elect.

OP 1103-6 RCP Operation 1107-3 Diesel Generator N SP 1100-3P ISI Misc. Valves 1010 T.S. Surveillance LP 1202-6A Loss of RC Coolant /RC Pressure OP Procedure Lock Valve List EP 1202-65 Loss of RC Coolant /RC Press with Auto HPI AP 1009 Station Organization & Chain of Commaand CP 1202-6C Loss of RC Coolant /with HPI, Core Fluod & LPI 1016 Operations Surveillance CP 1103-8 Approach to Criticality 1026/1407-1 Corrective Maintenance SP 1300-3Q ISI Test Valves !Jornal OP 1027 Preventat$ Maintenance SP 1300-3R ISI Test Inaccessible Valves AP 1013 By Pass of :afety Function & Jumper Control or 1104-4 Decay Heat Removat Solid System Operation Appendix to 1103-5 Rev. 12

$ SP 1301-1 Shift and Daily Checks AP 1004 Emergency Plan & Procedures AP 1014 Call Standby Personnel to Plant

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While thsra are many relisf end enfaty valvaa in sarvice ch2ra era only a few genaric typas which define the specifics of the test program.

Since existing facilities preclude full-scale testing at this k..') 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 sufficient 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 / l relief valve system tests, to be completed in July,1981. The  !

expanded valve test facility will be in place by July, 1981. I Scheduling of test facilities and other uncertainties could {

result in a longer schedule. Met-Ed believes, however, that i 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:

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" Program Plan for the Performance Verification of PWR Safety /

Relief Valves and Systems, December 13, 1979," forwarding letter updated.

10.3.3 Onsite Technical Support Center In response to the requirements of the Section 2.2.2.b, in NUREG-0578 as clarified by Mr. H. Denton's letter, dated October 30, 1979 to the licensee, Met-Ed has addressed clari-fications 1B, 1D, IE,1F and 1G by providing the following information:

Response to the clarification IB: The requirement to provide plans and procedures for engineering / management and staffing of the Onsite Technical Support Center would be contained in Emer-gency Plan Implementation Procedure (EPIP's). A preliminary EPIP titled Activation of the Technical Support Center (TSC) has been written. This procedure provides that the TSC be activated j for an Alert, Site Emergency, General Emergency, or as directed ^

by the Emergency Director. Emergency Actions required by this EPIP are:

a. Setting up of tables and chairs per the Technical Support Center Floor Plan, rx b. Arranging prints, technical manuals and other inform-(_) ation sources for availability.

10-6 Am. 15 i

c. Assigning a phonatalker to hook-up the operational line with the Emergency Control Center (ECC) to obtain in-formation on plant status, log calls, and advise the Energency Director when the TSC is operational.

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d. Moving the TSC personnel to the computer room, adjacent to the control room, in the event the TSC becomes unin-habitable, with the TSC Engineers reporting to the  !

Instrumentation and Control Shop. I Final operational conditions provided by the EPIP are:

a. The TSC would be operational with work areas set up for the Technical Support Staff and with communication lines established among the ECC, the Offsite Emergency Support Center and the TSC on the Operational Line.
b. The Technical Staff studying the emergency and plant status and making recommendations to the Emergency Direc-tor, via the communicator, to alleviate the emergency conditions and improve plant status.

Personnel reporting to the TSC are established by the TMI-l Emergency Plan Duty Section Schedule.

Response to clarification 1D: The TSC currently contains a Victoreen Vamp y -monitor and an AMS-3 continuous air monitor.

Both have alarm capability. In addition, the TSC is within the fg boundary of the Control Room Ventilation System, hence back up

(_) air monitoring is available. If any monitor indicates a signi-ficant increase, Health Physics would be notified and protective measures would be taken based on recommendations from the Health Physics Group.

Response to clarification IE: Records and drawings which describe the as-built conditions and layout of structures, systems and components have been assembled and will be placed in fire proof filing cabinets in the TSC. The required cabinets have been ob-tained and will be installed in near future.

Response to clarification 1F: With reference to the response to clarification IB, the EPIP described provides that the TSC be relocated to the computer room which has plant printouts, visual contact with the control room via a window, and would have voice communication with the control room. The TSC Engineers would use the nearby Instrumentation and Control Shop where room is avail-able to layout drawings, etc. and from which easy access is available to the control room and computer room when required.

Response to clarification 1G: As far as a longer range plan for upgrading the TSC is concerned, the following items have been considered to meet all the requirements.

f*s 10-7 Am. 15 1

Location

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The location of the TSC will not be changed and is described s ,/ in the Emergency Plan.

Physical Size and Staffing

.The TSC will house in excess of 25 persons, and contain neces-sary engineering data and information displays. Initial staffing levels to be accomplished within 60 minutes include engineers knowledgeable in the Reactor Core, Electric, Mechani-cal and Plant Computer Areas. Additional engineering personnel will be called in based upon the need for specific expertise as required.

Activation The TSC is activated in accordance with the requirements of NUREG-0610. Instrumentation in the TSC will be capable of providing displays of vital plant parameters from the time an accident begins.

Instrumentation The TSC will be able to access, display, and printout plant parameters independent from actions in the control room by use of Cathode Ray Tube Monitor System (CRT) with an attachad f- printer. This CRT system will be capable of accessing the

(_g) plant computer on a "not to interfere" basis. This same sys-tem will also be installed in the Of fsite Emergency Support Center (ESC).

Instrumentation Power Supply The plant computer is powered from a vital bus. The TSC CRT  !

System will be on a vital bus. A momentary loss of power to l either CRT system will not effect the plant computer, and  !

would only result in non-availability of the CRT during the  !

transient. No data would be lost during that period.

Technical Data '

All data currently available on the plant computer will be available via the CRT system. In any event the CRT system will be capable of accessing the information.

Historical data will be placed in fire proof filing cabinets in the TSC after the cabinets are installed. These cabinets l will contain a complete set of controlled as-built prints,  ;

P&ID's logic diagrams, wiring diagrams, and other types of I documents for use by the engineers in the TSC.

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Data Transmission

() Data available in the TSC will also be available in the off-site ESC.

Structural Integrity The TSC is totally within the Control Building in a safety related area. As a result, these requirements are exceeded.

Despite this, a back-up plan is available as noted in the response to clarification IB.

Habitability The TSC is totally within the Control Building and the Control Room Ventilation System; hence, the TSC meets the same stan-dards as the control room for shielding and ventilation.

Additional information is contained in the response to clari-fication ID on preceding pages. Permanent monitoring and dose reduction measures will also be provided.

On page C3-5 of the NRC's Status Report On The Evaluation of Licensee's Compliance, dated January 11, 1980, the licensee is required to address all the provisions of the proposed rule on emergency planning and is expected to report on those matters; and on any further upgrading of the Emergency Plan, in a supplement to this evaluation prior to any restart of

() TMI-1. The proposed rule change is to upgrade 10CFR Part 50 Appendix E to incorporate requirements from existing documents and to require NRC concurrence with state and local plans prior to issuing operating licenses or as a condition for licensee's to maintain operating licenses. A review is cur-rently underway on this proposed rule change. It is antic-ipated that the TMI-1 Emergency Plan provided to the NRC will not require changes to be in compliance. State and local emergency plans are being upgraded at the present time. The state plan has been reviewed and corrections are now underway to being the state and county plans into compliance.

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SUPPLDIDTI 1, PART 2 O ouxstron 2s E. Simulator Training The team concept for casualty control was stressed. The shift super-visor was evaluated in his command role.

. F. TMI Transient Constructive criticism of operator action during the transient was stressed in this portion of their training.

The elements of this program are being incorporated in the Shift Supervisors Development Program. Any person who will be subsequently assigned to the position of Shift Supervisor will be required to receive this training prior to the assumption of the duties and responsibilities of that position.

Position 4 Review Poliev The administrative duties of the Shifc Supervisor will be reviewed by ap-propriate members of the TMI Generation Group Staff in order to identify functions that detract fr a or are subordinate to the management respon-sibility for assuring safe operation of the plant.

O rw 1= ortw1 1 arec= =atic= 111d a=== =ca*7o-cember 31. Appropriate recommendations will be approved by the Senior i

Vice President - Mac-Ed, responsible for plant operations and will be implemented prior to March 1, 1980. Reviews such as this will be conducted on an annual basis in the future.

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SUPPLEMENT 1, PART 1 i.

QUESTION:

45. (Order Iteu. 1(d))

Your response to this item indicates that procedures have been.or are still being revised. Provide the procedures developed to define operator action during small break-LOCA's.

RESPONSE

EP 1202-6 has been provided for your review. This procedure defines operator guidance during small break LOCA. In addition, the B5W Guidance (B6W Document 69-1106001) used in developing the small break LOCA procedures and a summary of the supporting analysis (BSW Document 86-1105508-00) are attached. This guidance and its supporting analysis have been reviewed and are applicable to TMI-1.

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Documane No. 86-1105508-0 ATTACHMENT TO QUESTION 45, SUPPLDIENT 1, PART 1 l t

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ANALYSIS

SUMMARY

IN SUPPORT OF INADEQUATE CORE COOLING GUIDELINES FOR A

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O SUPPLEMENT 3 OPERATIONAL QUALITY ASSURANCE PLAN FOR THREE MILE ISLAND NUCLEAR STATION th UNIT 1 i

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TABLE OF CONTENTS O Statement of Policy and Autnority Introcuction 1.0 Organization 2.0 Quality Assurance Program 3.0 Control of Documents and Records 3.1 Instruction, Procedures and Drawings 3.2 Document Control 3.3 Quality Assurance Records 4.0 Design Control 5.0 Procurement and Material Control 5.1 Control of Procurement 5.2 Identification and Control o f Materials Parts and Components Control of Station Activities 6.0 6.1 Policy 6.2 Requirements 6.2.1 Details 6.2.1.1 Control of Inspection 6.2.1.2 Plant QA Monitoring 6.2.1.3 Control of Special Processes 6.2.1.4 Test Control IL 6.2.1.5 Control o f Measuring & Test Eouipment 6.2.1.6 Handling, Storace and Shipping 6.2.1.7 Inspection, Test & Operating Status p .

6.2.1.8 Fire Protection I

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6.2.1.9 Plant-Security

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6.2.I.11 Equipment Control 6.2.~1.12 Control of Cons truction, Maintenance (Preventive / Corrective), and j -Modifications I

6.1.1.13 Procedural Requirements 6.2.1.14 Control of Surveillance Testing and Inspection 6.2.1.15 Radiation Control 6.3 Responsibilities 7.0 Control of RGdioactive Wastes 1

8.0 . Control of Corrective Actions and Nonconformances 9.0 Audits Appendices

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() STATEMENT OF POLICY AND AUTHORITY It is the policy of the Metropolitan Edison Company to operate the Three Mile Island Nuclear Station so as to ensure the safety and health of the public and the person-nel on site.

It is also the policy of the Metropolitan Edison Com-pany to comply with the requirements of the Code of Federal Regulations, the NRC Operating Licenses and the applicable codes, guides and standards with respect to operation, inservice inspection, refueling, maintenance, procurement, repair and modificattan of the Station.

The President of Metropolitan Edison has delegated full responsibility for the Operation of TMI Unit 1 :o GPU Nuclear Corporation (GPUNC). This responsibility includes all aspects of operations, design, procurement and modification.

The President-GPU Nuclear Corporation has the respon-sibility and the authority to implement the Quality Assurance Program.

,s The Vice President-Nuclear Assurance reports directly

() to the President - GPUNC and provides, oy way of the Quality Assurance Department, the staff necessary to develop and maintain the Quality Assurance Program consis-tent with the applicable Federal and State requirements and to verify the implementation of the Program.

The Manager-Quality Assurance, who reports directly to the Vice Presioent-Nuclear Assurance, nas the overall authority and organizational freedom to identify quality assurance or management control proolems and provide recom.nended solutions. This authority and responsibility includes stop work authority in activities associated witn operations, maintenance, repair, modification, re fueling and manufacturing at or for the Three Mile Island Nuclear Station.

With regard to the stoppage of work, including the recommendation that an operating nuclear unit ce shut down, the Manager-Quality Assurance has direct access to -

the Vice President-TMI 1 and the Of fice of the President-GPUNC.

The Manager-Quality Assurance shall use this path when dif ferences of opinion regarding quality arise witnin GPUNC.

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The effectiveness of any Quality Assurance Program is 'l dependent upon the individuals who implement the program.

() Accordingly, all personnel of the General Puolic Utilles S9 stem and tneir contractors must comply with tne recuire-ments of this Quality Assurance Program. All memoers of management must give full support to maintaining an ef fective quality program as defined in this Plan.

The-Quality Assurance Program, as descrioed in this Plan, is approved for implementation at Three Mile Island Unit 1.

Date President, Met-Eo Date Pres 10ent, GPU Nuclear Corp.

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