Text

Radiation Protection Plan,App 7A,Revision 3
	ML19323G343
Person / Time
Site:	Three Mile Island
Issue date:	05/20/1980
From:	Herbein J, Potts W; METROPOLITAN EDISON CO.
To:	;
Shared Package
ML19323G337	List: ML19323G336; ML19323G340; ML19323G343; ML19323G345;
References
	PROC-800520, NUDOCS 8006020205
	Download: ML19323G343 (100)
	v • d • e

R. vision 3 APPENDIX 7A May 20,1980

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THREE MILE ISLAND NUCLEAR STATION UNIT 1 RADIATION PROTECTION PLAN r

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/

/ W. E. Potts Manager, Radiological Controls l\ s l

\

. G. Herbein Vice President Generation, Met-Ed l

l Changes to this document require approval by these positions.

9 8006030205

4

. Rnvision'3 May.20, 1980 Article 7 - Control of Radioactive Contamination O Radioactive surface contamination shall be controlled in order. to minimize possible inhalation or ingestion of radioactivity and to mini-mize build up of radioactivity in the environment. Measures to contain radioactivity and to minimize the number and extent of areas contaminated shall be taken in order to minimize personnel radiation exposure, to simplify subsequent personnel and area or facility decontamination, and to minimize the need to rely on anticontamination clothing.

The release for uncontrolled use surface contamination level and skin contamination action level for beta-gamma activity shall be 1000 dpm/100cm 2 for smearable contamination and 0.lmR/hr for total contamination. The preferable measurement technique is with a pancake frisker. For alpha activity, the surface contamination level for uncontrolled release and skin contamination action level is 100 dpm/100cm2, Emphasis in planning, training and working shall be placed on mini-mizing the numbers of occurrences and amounts of radioactivity involved in occurrences of radioactive surface contamination of a person's skin or of areas not controlled for radioactive surface contamination. Each such occurrence shall be reviewed in detail to determine how to correct deficiencies and improve control of radioactivity.

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TABLE 8A-9 LOSS OF NORMAL FEEDWATER CASE NO. 14 Makeup: Available Letdown: 1solated at T = 5 see EFW Mech: Available EFW Steam: Available EFW Capacity: 500 gpm Atmos Dump: Available Will open at quick open setpoint since condenser available Turbine Bypass: Available; Setpoint = 1025 psia Small Safeties: 1060/986 psia Bank 1: 1065/990 psia Banks 2&3: 1070/995 & 1072/997 psia.

Press 11trs: 5 Banks ]

(sm)g Press Spray: Available on T = 0; Off on Rx Trip RC Pumps: Available RPS Trip: High Pressure (2405 psia)

RPS Defeated: Turbine Trip SFAS Trip: 1600 psig RCS 4 psig Contain Press Defeat Diesel Generators: 2 Offsite Power: Available Decay Heat: 1.2 x ANS ,

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Recommendation Response To be Submitted
() 7. Financial Qualifications TMI-2 Lessons Learned Recom-Separately 8.
mendations - NUREG 0578 2.1.1 Section 2.1.1.3 and S1 P1 Q11
& 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, 92, 93, 94 & 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 Section 2.1.1.7 & 2.1.2.6 2.1.7 r' 2.1.8a Section 2.1.2.4 2.1.8b Section 2.1.2.7.2 2.1.8c Section 7.3.5.3 2.1.9 Sections 3.1.1, 6.0, 8.1 2.2.1.a S1 P1 Q40 aad S1 P2 Q25 2.2.1.b Section 5.0, S1 P1 Q42 and S1 P2 Q27 2.2.1.c Section 3.0, S1 P1 Q43, S1 P1 Q28 and Q29 2.2.2 Section 4.0, Section 10.3.3, 10.3.4 and S1 P1 Q44 2.2.3 Not Applicable until NRC Regulations are revised
() 10-3 Am. 1{
Data Transmission Data available in the TSC will also be available in the of f-site ESC.
Structural Integrity The TSC is totelly within the Control Building in a saf ety related area. As a result, these requirements are exceeded.
Despite this, a back-up plar, is available as noted in the response to clarification ib.
Habitability The TSC is totally within the Control Building and the Control i;oom Ventilation System; hence, the TSC meets the same stan-dards as the control room for shielding and ventilation.
Additional inf ormation is contained in the response to clari-fication ID on preceding pages. Pe rmanent 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, i () in a supplement to this evaluation prior to any restart of TMI-1. The proposed rule daange is to upgrade 10CFR Part 50 1 Appendix E to incorporate requirements frva existing documents and to require NRC concurrence with state and local plans prior to issuing operating licenses or as a condition for li ce ns ee 's to maintain operating licenses. A review is cur-rently unde rway on this proposed rule change. It is antic-ipated that the TM1-1 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 nma underway to being the state and county plans into compliance.
10.3.4 Onsite Operational Support Center Met-Ed has addressed Section 2.2.2.c of NUREG-0578 in Section 4 7.1.3. The following is additional information regarding communication and management of the OSC.
Communication bebeeen the Emergency Control Center (ECC) and the OSC is available via a maintenance and instrumentation phone line. This is hardwired and in place. It is activated by plugging in the head sets at each end. A phone talker is assigned to each end of the line. Back-up is provided by the Public Address System and the Emergency Radio System.
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10-9 Am. 18 l
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Requirements for communicating . logging, etc. are to be provided by an EP1P titled: Communications and Recordkeeping which also
() contains log sheets for each communicator.
Management of the OSC is provided by the Operations Support Center Coordinator who is assigned by the TMI-l Emergency Plan Duty Section Schedule. From page 5-9 on the Emergency Plan, these duties are described as follows: "(2) The Operations Support Center Coordinator shall be responsible f or and direct the activities of personnel reporting to the Operations Support Center. In additien, he will also direct and coordinate the activities of the Health Physics and Chemistry Coordinators.
As mentioned above, he will dispatch of f site and/or onsite Radiological Monitoring Teams as directed by the Radiological i Asses sme nt Coordinator."
10.4 Transient Analysis and Procedures for Management of Small Breaks As part of the NRC Order of August 9, 1979, long term actions recommended by the Director of Nuclear Reactor Regulation were included. Item two of that list recommended that the Licensee:
" . . . give continued attentica to t ransient analysis and procedures for management of small breaks by a formal program set up to assure timely action of these matters;"
As part of the development of the TMI Generation Group, a new O section was established as part of the Systems Engineering Depa rt me nt. This section, known as the Plant Analysis Section, is charged with responsibility for a continuing review of plant pe rf ormance. This will include an on going technical evaluation of the overall plant performance as well as the performance of major systens and components.
The Plant Analysis Sect ion is also charged with the responsi-bilities of reviewing key information f rom other nuclear plants.
This inf ormation will be obtained f rom the review of Licensee Event Reports, general industry survey information, indu s t ry -
contacts, regulatory body documents, owner's group activities, and s tanda rds committees. Seve ral of these sources of inf or-mation will be provided to the Plaat Analysis Section by other functional sections within the TMI Generation Group. The ove r-all industry approach to the interchange of such inf ormation is currently under development and is expected to utilize organizations such as the Nuclear Safety Analysis Center and the Institute for Nuclear Power Operations. GPU fully intends to participate in these information interchange activities, when developed.
Recommendations resulting f rom these raviews may be used to modify equipment design, operating and maintenance methods,
- operator training programs, procedures, or other aspects of
(,j plant operation.
10-10 '
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Additionally, two other sections with the Systems Engineering Department provide more specialized analytical functions that,
{-3/ on a continuing basis, improve understanding of transient analysis and procedures for the management of small breaks.
The Control and Safety Analysis Section has the in-house capability to perform analysis of the type shown in Section 8 of this report.
They also contract for and review the results of such work performed by the NSS supplier or other consultants. The Control and Safety Analysis Section has been deeply involved in analysis of programs for small break LOCA operator guidelines and anticipated transient operator guidelines. This work is closely associated with and has had a direct input to operator training programs and procedures.
The Nuclear Analysis Section performs or reviews fuel and core related aspects of transient and accident analysis, including reload evaluations. This work is interf aced with the activities in the Control and Safety Analysis Section.
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SUPPLEMENT 1, PART 1 O QUESTION:
10i. Verify that at least one EFW system pump including its associated auxiliaries such as the turbine driven pump lube oil cooling system, and its associated flow path and essential instrumentation will automatically initiate EFW system flow and is capable of being operated independently of any alternating current power source for at least two hours. Conversion of direct current power to alternating current is acceptable.
4
RESPONSE
The Babcock 6 Wilcox Company has prepared a report for GPU Service Corporation l
on the reliability of the steam driven EFW pump and its auxiliaries. The report, entitled, " Emergency Feedwater System Reliability Analysis for Three Mile Island Nuclear Generating Station, Unit 1, Revision 1" was issued in December 1979 and submitted to NRC by letter dated February 7, 1980 (E6L-2102).
In addition, GPUSC reviewed the adequacy of the steam supply to the turbine driven emergency feedwater pump. The B6W reliability analysis was based on the assumption that with a degraded steam supply (i .e. , MS-V6 failed open),
the turbine-driv"n pump would start and continue to run. The summary of the results of our review of this assumption re as follows.
In the initial plant design, the possibility of a failure of the pressure control valve, MS-V6, was considered. To cope with a failure of this valve to control pressure, safety valves (MS-V22A/B) were included to protect the EFW steam supply line and the turbine steam chest from overpressurization.
As indicated in the B6W reliability report these valves were set at 505 and 495 psig. Recently the turbine vendor, Worthington, was requested to comment on the ability of the turbine to operate with inlet stcam pr::: ares as high as 500 psig. Worthington noted that the turbine governor and trip valve were only rated at 250 psig. Therefore, as the design currently exists today, the turbine driven EFW pump cannot be expected to operate if pressure control from MS-V6 fails open and the results of the BGW reliability report are invalid.
Met-Ed has reviewed the situation and determined that the above stated problem can be resolved by resetting MS-V22A/B to approximately 250 psig. There fore ,
prior to restart, MS-V22A/B will be returned to the vendor for spring replacement and resetting. This action will protect the governor and trip valve from damage due to overpressurization, and will re-establish the validity of the B6W reliability analysis. In addition, it is our intention to add an air bottle to the air supply for MS-V6 to assure operation of this valve on loss of AC power, as discussed in response to Question 10j in Supplement 1, Part 1. This action is desirable to avoid the continuous cycling of the relief valves under the above assumed condition and set pressures.
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(_j In addition, a review has been made to ensure there are no other dependencies which could prevent the EF pump turbine from operating when needed.
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. SUPPLEMENT 1, PART 1 RESPONSE TO QUESTION 101 PAGE 2 O
1 The emergency feed pump turbine bearings are lubricated solely by ring lubrication. The oil ring picks up oil from the reservoir in the bearing case and carries it up to the shaft. Integral wster jackets in the bearing cases are used to cool the oil. Cooling water is provided from the pump discharge through a selt regulating control valve with the water being returned to the condensate storage tanks. All bearing cooling water piping is Seismic Class I.
The bearings of all three emergency feed pumps are lubricated with constant level oilers which do not require pumping. The oil is cooled by a water jacket around the oil reservoir. Water is supplied from the pump discharge back to the condensate .-torage tanks, similar to the turbine bearing cooling.
Again, all piping is Seismic Class I and requires no external power or control.
Therefore, with the addition of the air bottle for control of MS-V-6 it can be seen that there are no AC dependencies which could prevent emergency feedwater from being initiated upon loss of all AC power, as supported by the B6W analysis.
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SUPPLEMENT 1, PART 1 f-') QUESTION:
'O 10j. We require that you evaluate the consequences of a postulat-ed break in the steam line to the turbine-driven EFW pump to determ*nt the need to qualify the EFW system valves, valve actuators, and instrumentation for the environmental conditions resulting from such a high energy line break in order to maintain operability of the motor-driven EFW pumps and their associated flow trains.
RESPONSE
The consequences of a postulated break in the steam line to the turbine driven emergency feedwater pump have been evaluated in Supplement 2, Part IX, Amendment 41 (7/16/73) of the TMI-l FSAR including arrangement drawings for the system.
The following is a summary of the evaluation contained in Part IX of Supple-ment 2:
Inservice inspection of certain specified break locations will provide assurance that rupture will not occur at those respective points.
The primary means of effecting a cooldown after a postulated break p(,s,2. outside containment would be the emergency feedwater system with high pressure / low pressure injection cooldown serving as backup.
The main steam piping system has been conservatively designed and analyzed in accordance with B31.1.0 Code, For Power Piping.
The main steam piping welds were 100% radiographed from the steam generators up to the turbine stop valves.
The main steam supply to the emergency feedwater pump turbine piping welds were 100% radiographed.
Additionally, the motors (2) for the motor driven emergency feedwater pumps have been certified by the manufacturer to start and remain operable even under the highly unlikely postulated event of a main steam line break in the auxiliary building. The environmental conditions resulting from the postulated main steam line break and during which the motors will start and remain operational are 323*F returning to ambient in approximately one hour.
The associated valves (COV10A & B; EFVlA & B) in the motor driven emergency feedwater system are normally open and are not required to change direction (position) for the system to become operational. EFV30A & B are under the control of the ICS and respond to SG level requirements.
() s-There is no instrumentation required to place the motor driven emergency feedwater system into operation. See also Supplement 1, Part I response to Question 101 and Supplement 1, Part 2 response to Question 14.
Am. 18
SUPPLEMENT 1, PART 1
(~
QUESTION:
29. You state (Section 10.2) that Item 1 of Bulletin 05A, which requires you to review a chronology of the TMI-2 accident so that an understanding of the events will ensure against such an occurrence at Unit 1, as "not applicable". Correct this statement and describe the adequacy of your review.
RESPONSE
Following the TMI-2 accident an ad hoc committee was formed to review the accider.t and recommend appropriate modifications to be implemented at TMI-l in light of the " lessons learned". This ad hoc committee consisted of management, operations and engineering personnel who were involved in the accident effort. Most af the recommendations of this committee were included in Met-Ed's letter dated June 28, 1979, some of which were later superseded by similar recommendations of the NRC TMI-2 Lessons Learned Task Force (NUREG-0578).
In addition to the above, an Accident Investigation Task Force was established g(_g
/ to analyze the TMI-2 accident in depth in order to completely understand the event that occurred and the reasons they occurred. Evidence of our understand-ing of the accident is provided by our quarterly progress reports for TMI-2 recovery (DiI-II-RR-6) .
An implementation committee has also been recently formed to review the implementation of the lessons learned and verify that they are being accomplished in a manner consistent with the ad hoc committees (see above) recommenda tions.
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Am. 18
SUPPLEMENT 1, PART 1 QUESTION:
42. Paragraph 5.4.8 of the Met-Ed/GPU TM1-1 Restart submittal outlines the educational background requirements for a Shift Technical Engineer (STE).
a. Provide the details of the training the STE will receive to assure a thorough knowledge of normal reactor operations, anticipated transients, and effects of multiple equipment failures and operational errors.
RESPONSE
The Shift Technical Advisor (STA) Training Program consists of two parts.
Part 1 includes TMI-1 plant specific material and is to be completed by January 1, 1981. Part 2 includes college level courses and is to be completed by January 1, 1982. The STA's are all holders of Bachelor of Science degrees in engineering or related science areas. Part 2 of their training program will serve to supplement their previous academic endeavors. An outline of the STA Training Program follows:
PART 1 O
\- / I. Plant Systems A. Primary
1. Reactor Coolant System
2. Decay Heat Removal
3. Makeup and Purification System
4. Intermediate Closed Cooling System
5. Reactor Coolant Pumps
6. Pressurizer Level Control System
7. Pressurizer Pressure Control System
8. Rod Control System
9. Liquid Radwaste System
10. Gaseous Radwaste System
11. Once Through Steam Generator
12. Reactor Vessel and Internals
13. Spent Fuci Cooling
14. Steam Generator Feedwater Control B. Secondary
1. Condensate
2. Feedwater
3. Condenser Air Removal
4. Condenser Circulating Water f3 V S. Turbine Bypass (Steam Dump)
6. Instrument Air
7. Main Steam
8. Emergency Feedwater
9. Auxiliary Steam Am. 18
SUPPLEMENT 1, PART 1 RESPONSE TO QUESTION 42 PAGE 2 O
C. Auxiliary
1. Plant Ventilation
2. Radiation Monitoring
3. Seismic bbnitoring
4. Pump Noise Monitoring D. Engineered Safeguards
1. Engineered Safeguards Actuation Systems
2. liydrogen Purge
3. liydrogen Recombiner E. Instrumentation and Control
1. Integrated Control System
2. Electro-llydraulic Control System
3. Nuclear Instrumentation (Out of Core)
4. Incore Detectors
5. Reactor Protection System
6. Control Rod Drive / Indication F. Electrical
1. Emergency Diesel
2. Vital Power
3. Main Distribution
4. Turbine Generator i II. Procedures and Documentation i
A. Procedures
1. Emergency
2. Operating
3. Abnormal Operating
4. Surveillance
5. Ilealth Physics
6. Administrative B. Documentation
1. License (Technical Specifications) l
2. Regulations (10 CFR 19, 20, 50, 55) l
3. Plans (QA, Fire Protection, Radiation, Emergency)
4. Power Reactor Events
5. Switching and Tagging 1
)
Am. 18 1
I
SUPPLEMENT 1, PART 1 RESPONSE TO QUESTION 42 PAGE 3
(~
6. Shift Turnover / Logs
7. Work Requests
8. Change Mods
9. Document Control
10. FSAR III. Transient and Accident Response A. Accident Analysis
1. LOCA - Large and Small Break
2. Reactor Coolant Inventory Increasing Events
3. Reactivity Addition Accidents
4. Core Overheating Transients S. Core Overcooling Transients
6. Reactor Coolant Flow Transients B. Unusual Event Recognition and Response
1. ICS Transients
2. Plant Transients
() C. BfW 4 Simulator Training
1. Control Room Familiarization
2. Reactor Startup/ Shutdown
3. Plant lleatup/Cooldown
4. Plant Operation with ICS S. Transient and Accident Response IV. Related Operator'.caining A. Nuclear Power Fundamentals B. Operator Accelerated Retraining Program (OARP) Modules 2 through 6 C. Ilealth Physics V. General Supervisory and Career Development A. Supervisor Development Program
1. Decision Analysis
2. Probability Assessment i
3. Team Development
4. Command Boundary Conditions rN S. Practical Exercises O
Am. 18
SUPPLEMENT 1, PART 1 RESPONSE TO QUESTION 42 PAGE 4 4
O PART 2
1. General Technical Education A. Math B. Chemistry C. Physics / Reactor Physics D. Thermodynamics / Heat Transfer E. Control Systems / Electrical Fundamentals F. Fluids G. Communications
11. Materials T Because of the six shift rotation (one of which is a training shift), this
(~/
\- program represents approximetely 640 hours0.00741 days 0.178 hours 0.00106 weeks 2.4352e-4 months of training over the two year initial training period. This program has not been finally approved and is still subject to revision.
i Am. 18 m
SUPPLEMENT 1 - PAR
2_
RESPONSE TO QUESTION 3 PAGE 2
,V V
lower flowrates might also be acceptabla in meeting the defined acceptance critaria.*
Note that while the RC pump heat could not be accounted for diractly in the attached CADDS analysis, hand calculations were done to confirm that inclu-sion of the pump heat would allow the acceptance criteria (pressurizer does These hand calculations super-not fill solid) for the LOW event to be met.
imposed the RC pump heat on the CADDS analysis.
One assumption was not specifically addressed in the analysis was worst case OTSG pressure. Heat removal is controued by steam pressure; a higher steam The highest pres-pressure will re=ove less heat than a low steam pressure.
sura possible in the steam lines is the safety valve set pressure (1050 psi).
The heat removal capability of energency feedvatar =ay be represented in terms of energy removal versus time by taking the product of the flowrate and the feedvater enthalpy rise through the generator. For this analysis, the feedwater temperature at the entrance to the generator was assu=ed to be 100F, with a corresponding enthalpy of 71 Btu /lbm. The enthalpy of saturated steam does not change appreciably between 900 psia and H00 psia (u96 Stu/
lbm to H89 Beu/lbm) . This indicates that the effect of pressure change on enthalpy diffarance is small. The transient s4-nlation assumed that after initial fluctuations, the staan pressure rema4,ed at 1015 psig. Review of U
(') plant post-trip data indicates that this is a realistic value. The differ-enca between 1015 psig (1192 Stu/lbm) and a possible worst case of approx 1-mately 1050 psig (1150.5 Bru/lbm), based on safety valve setpoints, is insignificant and would not impact the analytical conclusion regarding the adequacy of 500 gpm.
It should also be noted that this analysis was for the purpose of showing the adequacy of 500 gym and not to determine the mini um AW acceptable for the transient. Therefore, it is possible a lower flowrate would produce acceptable results given the same assumptions and acceptance criteria.
The events which require emergency feedwater are:
1) Loss of Coolant Flow
2) Loss of Offsite Power (LOOP)
3) Main Stamm14ne Break (MSL3)
4) Small Break LOCA's (Certain sires)
5) Loss of Feedwater (LOW)
The most demanding event in terms of the need for heat removal via EW is the LOW without LOOP since this event requires the removal of RC pump heat as vall as decay haat. Therefore, based on the above discussion, the 500 pgm EW flow availabla under single f ailure assumptions is adequate
/7 *The acceptance critaria for the m4 4 "" anH 14=ry flow rate were that (1)
[ the pressurizer does not go solid and (2) the elaccromatic relaaf valve does not actuata. An an* 14=ry flow rata of 500 gym was found to seat
~
these critaria. The assumptions used for these analysis are conservative for DiI Unit #1. (720 gpm is available under non-singic failure conditions).
Am. 18
SUPPLEMENT 1, PART 2 RESPONSE TO QUESTION 3 PAGE 3 d(,
Main Feedline Break is a somewhat more abrupt case of LOFW. However, from a long term cooling standpoint, the heat removal requirements are identical to those occurring during a LOFW. Therefore, although the initial response to a feedline break may be more severe, the emergency feedwater sizing requirements based on LOFW considerations es ;ures sufficient heat removal capacity to mitigate the line break accident . Therefore, the results of a LOFW and Feedline Break Accident are essentially the same.
A review of RCS pressure history during various trip settings (and times)*
i shows that peak RCS pressure occurs before the application of EFW (i.e. , in less than 40 sec.). Although the attached analysis assumes a reactor trip on anticipatory trip, the 500 gpm EFW flow requirement is supplied after 40 seconds and is not a factor for peak RCS pressure in the short term.
Long term peak pressurc is influenced by EFW flow and the attached analysis demonstrates that 500 gpm is satisfactory and does not result in safety valves lifting even assuming a 1.2 multiplier on ANS decay heat.
l l
l l
l
Letter - Met-Ed (J. G. Herbein) to USNRC (R. W. Reid), on "High Pressure Trip and Pressurizer Code Safety Valve Settings", GQL-0669, April 17,1978.
(This analysis was done to justify a reactor trip setting of 2390 psig and the post restart setting will be 2300 psig).
i i
Am. 18 l
I I 4
i i
+
j 3UPPLEMENT 1, PART 2 [
l O i I QUESTION f f (
6. Provide design drawings for the modifications which provide for i control room annunciation of all automatic start conditions of ,
the EFW system.
i !
! RESPONSE l
l The design for the modifications which provide for control room annunciation
{
of all automatic start conditions of the EFW System, are shown in drawings l SS-209-755, Rev. IC-1 and SS-209-iS6, Rev. IC-1. The attached Table 6-1 i l lists the instrumerts used in these drawings and their respective functions. l i
)
j
!O
)
i i
i t
1 i
1 1
4 O
{
I l
Am. 18
SUPPLEMENT 1, PART 2 4
[3 ATTAClBIENT TO QUESTION 6 m) b TABLE 6-1 Drawing No. Instrument / Device Service SS-209-755 DPS-829 FW-P1A Differ. Pres:5.
(Train "A") DPS-830 FW-P1A Differ. Press.
DPS-938 SG."A" FW-MS Differ.
Press.
DPS-939 SG."B" FW-MS Differ.
Press.
62 x 1/ Rack A(4) RC Pumps Power Monitors O
V SS-209-756 DPS-542 DPS-543 FW-P1A Differ. Presa.
FW-PIE Differ. Pres..
(Train "B")
DPS-940 SG."A" FW-MS Differ.
Press.
DPS-941 SG."B" FW-MS Differ.
Press.
62 x 2/ Rack B(4) RC Pumps Power Monitors t
\./
Am. 18
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es 1 emame-
SUPPLEMENT 1, PART 2 QUESTION
14. Your response to question 10j provided in Amendment S is not complete.
Provide legible arrangement drawings for the EFW system showing the location of all system pumps, piping and valves. Provide qualifi-cation documentation which assures that the motor driven EFW pumps will start and remain operational under the environmental conditions (humidity and temperature) resulting from a postulated break in the main steam supply line to the turbine driven EFW pumps. Further, verify that the EFW control valves and actuators are qualified to function under these environmental conditions. Also, provide an analysis which justifies the environmental conditions (323 F) assumed as a result of the postulated steam line break.
RESPONSE
As d3 scribed in response to Question 10j (Supplement 1, part 1) the subject break was not considered probable enough to warrant detailed design consider-ation at the time TMI-1 was licensed. Since TMI-1 was licensed, NRC acceptance criteria for EFW systems has been modified and the EFW system has taken on new importance. In recognition of this fact, Met-Ed has initiated a complete design review of the EFW system to upgrade it to the current licensing criteria to the extent practicable. This review will consider and resolve the type of gs concerns raised by questions 12, 13 and 14 above. We believe that this approach
(_) is preferred over an item resolution of issues. Nevertheless, a response to your specific concerns is given below.
Arrangement drawings showing the location of important EFN valves and piping was provided separately on November 28,1979 (
Reference:
E-304-086, Rev. 14).
The EFW pumps have been certified to withstand the calculated environment.
A copy of the motor qualification certification is attached, together with the calculations which support the environmental conditions (323*F). Environ-mental qualification of EF-V30A/B to 323*F was not invoked as part of the original purchase order for these valves. We have determined, after extensive review of the valve operators with the vendor, that certain elements of the operator cannot withstand the necessary accident environment.
In order to improve the reliability of the EFW, the EFW control valves (EF-V30A and EF-V30B) and the EFW turbine throttle valve (MS-V6) will be upgraded. The effort to upgrade these critical EFW valves involves a qualifi-cation program with seismic and environmental aspects. The qualification program will serve to identify valve components which would not perform properly in the expected accident environment. The subsequent application of analytic techniques will demonstrate that the modified EFW valves will perform properly under the postulated environmental and seismic conditions. This effort _ including necessary hardware replacement will be completed before restart.
I h
V Am. 18
SUPPLEMENT 1, PART 2 RESPONSE TO QUESTION 14 PAGE 2 l')
\/
In addition to changes associated with the environmental qualification progrsm for EFW valves EF-V30A, EF-V30B and MS-V6, changes to the control systems for these valves will be made to assure reliable operation of these valves under loss of AC power operation. These air operated valves, which are required for EFW operation via the EFW steam turbine pump, receive their motive power via an AC powered air supply system and thus would not operate under loss of AC power conditions. Figure 1 (attached) shows the modified control design for EFW Valves EF-V30A and EF-V30B. A two hour air supply is shown in the upper right corner of Figure 1 which would allow operation of EF-V30A/B independent of the AC air supply system. The two hour air supply (via a 3600 psi nitrogen tank) is located in a non-hostile environment and thus need not be environmentally qualified. An additional modification calls for replacing the existing E/P converter (voltage to pressure signal converter) with an E/I (voltage to current converter) and I/P (current to pressure converter). This modification was required since an environmentally qualified E/P converter could not be obtained. The replacement I/P converter will be environmentally qualified; the associated E/I converter is located in a non-hostile environment (the Control Building) . Figure 2 (attached) the control design for EFW Turbine Control Valve MS-V6, shows the same back-up air supply system as used for EF-V30A/B.
O o
Am. 18 4
SUPPLEMENT 1, PART 2
, ATTACINENT TO QUESTION 14 l
\
I a 18 VOC SIGNAL FROM ICS BONNDSTILE ENVIROEMENT l RELIEF ONLV NO CERTIFICATION
!_ _E4 gy 4 20 MAOC TO COETROL ADV1 E, REQUIRED
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P051T10hER l FEED BACK
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'#^ EF V30AlB e gk ggp ,
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l VOLUME 8 i TANK I l
, i .
l l t_ TEE y[u _ _ _ ,j '
- . . _ _ .J FIGURE 1 MODIFIED CONTROL SYSTEM FOR EFW VALVES EF-V30A AND EF-V30B i
l l O l V l
Am. 18
)
SUPPLEMENT 1, PART 2 ;
ATTACINENT TO QUESTION 14 m
U S$f A.S. FROM CONTROL 4 q . _ ___ .__ g,3,ggppty SCMEME 1
, L I r*
>> s ,, ,, l l -( L. __
e i
..E, , EFPT C EF-Pt EF . SYSTEM PREFIX " EMERGENCY FEEDWAl!P*
d EFFT EMERG. FEED PUMP TURBINE g <>
MS SYSTEM PAEFIX " MAIN STEAM" O
FIGURE 2 i
MODIFIED CONTROL SYSTEM FOR FOR EFW TURBINE CONTROL VALVE MS-V6 O
Am. ~ 18
SUPPLEMENT 1, PART 1 7_ QUESTION:
!! 21. (Restart Submittal Section 2.1.1.5.1.3) explicitly state whether or not the non-ECCS support services for RCP operation will be upgraded to Seismic Category I AND protected from BOTH pipe whip and jet impingement.
RESPONSE
Intermediate cooling and nuclear services cooling water is designed and installed Seismic I. We proceeded to analyze these piping systems inside l containment for pipe whip, jet impingement and gravity missile hazards in accordinace with Regulatory Guide 1,46 and Standard Review Plan 3.6.2. This analysis is now complete. The analysis shows that these piping systems could not be protected under those conditions. Consequently, a rupture detection system will be provided to meet the following criteria:
1. Leak detection and isolation protection are to be provided for Reactor Building pressure conditions which are below 30 psig such that if there were a leak in the piping system inside of the Reactor Building during an accident, then that leak would be into rather than out of the Reactor Building.
2. The system Icakage rates should be detected and alarmed
(^) (in the control room) . If the leakage rates are undetected,
\/ they would reduce the surge tank volume from high to low level within one hour or less.
3. The containment isolation valves should be closed when the level in the surge tank reaches the low-low level only if a safety injection (HPI) signal had been initiated.
~
Reactor Building normal- cooling and primary shield cooling will maintain the current containment isolation on 4 psi in the reactor building and will be modified to provide diverse containment isolation of the cooling water system based on high pressure injection initation at 1600 psig.
/^
_ ,/
Am. 18
SUPPLEMENT 1, PART 2 RESPONSE TO QUESTION 25, PAGE 2
~
ks)
F E. Simulator Training The team concept for casualty control was stressed. The shift supervisor 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 4 responsibilities of that position.
Position 4 Review Policy l'
() The administrative duties of the Shift Supervisor have been reviewed by appropriate members of the TMI Generation Group Staff in order to identify functions that detract from or are subordinate to the management responsibility for assuring safe operation of the plant. l (See attached memorandum.) No changes were determined to be necessary.
In addition, reviews such as this will be conducted on an annual basis in the future.
l i
(_)
Am. 18
. SUPPLEMENT 1, PART 2 ATTACIDlENT TO QUESTION 2S j
/AETROPOLITAN EDISON CO M P A N Y s. mi,2rv of caers w,c viun corporai on Loc: tion TMI Mulcear Station Subject SHI T SUPERVISORS DUTIES Middletown, Pa. 17057 REF: NU REG 0573 Section 2.2.la.
Date December 10, 1979 To G.P. Miller J.C. Herbein NRC Nu Reg. 0578 short term recommendations require that the administrative duties of the Shif t Supervisor be reviewed by the senior of ficer of each Utility.
In that his Technical and Safety responsibilities have already been d'e fined by our senior management only the administrative duties of the Shift Supervisor will be addressed.
Listed below for your review and further action are the administrative duties of the Shift Supervisor:
ITEM ALTERNATIVE RECOMMENDATION A. Log Books Maint ained None Leave As Is none B. Checklists Reviewed or O
V Initiated
1. Shift Supervisor Turnover None-required by Leave As Is
2. ESAS Checklist initiated by Nu Reg.
CR0 reviewed by Shift Supervisor None-required by Leave As Is Nu Reg.
C. Coordination Activities Coordinates Maintenance, HP, None-In that these duties Leave As Is Security events (but does not do not distract from directly supervise) that h2s primary safety re-occur on his shift. lated duties and en-hance plant safety by ensuring Senior Licenced Personnel are aware of items that could affect plant operation and public safety.
D. Paper Work Reviewed Uhile on_
Shift _ Leave As Is
l. Reviews deficiency sheets None-NRC commitment to generated during Performance have Senior on site of Tech Spec surveillances person review-im-mediately. This in-creases plant safety and goes along with his
/]
D primary function of plant safety. l
}
l l
INTER-OFFICE MEMORANDUM l
G.P. }! iller-J.G.'Herbein page 2 d
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ITEM ALTERNATIVE RECOM'tE.JATICE G
2. S'igns plant access lists Infrequent item. Does Leave As Is required during his shift not occur frequei.t enough to be of a burden or distraction.
Alternative would be to have Superintendent called.
3. Signs RUPs/ Work Requests This is done only as Leave As Is and Switching and Tagging his option. This is the Requests primary duty of the Shift Foreman. He has no require-ment to do this.
4. Signs and reviews special Alternative is to have the Leave As Is operating procedures- Shift Foreman sign, hewever this is a safety related function and one of the primary functions of the Shif t Supervisor.
5. Signs and reviews liquid / None- This is one of his Leave As Is gaseous releases primary functions and goes along with his duties of ensuring
("~) Public Safety.
6. Reviews Circular / Bulletins None-In that this broadens Leave As Is assigned to him by the his knowledge of Plant Supervisor of Operations and Public Safety and is donc at his con-vience,no change is necessary.
None-This item keeps him Leave As is
7. Review and admends Operating and Emergency current and is done Procedures assigned to him at his convenience.No by the Supervisor of change is necessary Operations It should be noted routine duties such as union time sheets and neal slips are the responsibility of the Shift Foreman.
The Administrative duties of the Shift Supervisor are consistemt with the Un Reg requirements and do not require any deletions or additions. It is requested these items he reviewed with our senior management personnel and a letter containing their comments be generated to document their review.
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M.J. ROSS I
SUPERVISOR OF OPERATIONS-C;IT I i
FUR:dly cc: C. Broughton (GPU)
J. Colitz (GPU)
R. Harding (Licensing)
D. Slear (GPU)
P. Walsh (GPU) 1' File: FUR's Writets File i l
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SUPPLEMENT 1, PART 2 RESPONSE TO QUESTION 52, PAGE 3
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In addition to the modifications described above, a ventilation system to mitigate the consequences of a postulated fuel handling accident in the FHB will be installed. This new system will meet the requirements of Regulatory Guide 1.52, Revision 2. This system, and intermediate modifications to the Auxiliary and Fuel Handling Building Ventilation System, are described
< below.
The Auxiliary and Fuel Handling Building Ventilation System will undergo extensive modifications which will be undertaken in two phases as described below.
Phase 1 Prior to restart, the TMI-1 ventilation system will have been modified as shown in Attachment 2. The follow.ng equipment will have been added:
1. Damper U with interconnecting ductwork.
2. Damper T and Fan F with interconnecting ductwork.
In the event that contamination (radioactivity) is sensed in the ductwork, r3 Dampers U and T will close, Fan F will trip, and ventilation will be via
\,
) filter trains M and N.
Phase 2 Prior to the next refueling, the following additional modifications will be made, as shown in Attachment 3:
1. Filter trains Q and R will b.e added. From the intake side of the filter trains, the composition of the train will consist of a prefilter, an electric heater, a HEPA filter, a charcoal filter, and a final HEPA filter. The design of filter trains Q and R will meet the requirements of NRC's Regulatory Guide 1.52, " Design, Testing and Maintenance Criteria for Post Accident Engineered-Safety Feature Atmospheric Cleanup System Air Filtration and Absorption Units of Light Water-Cooled Nuclear Power Plants".
2. Dampers S and V will be added.
3. Fans G and H will be installed together with all connecting ductwork and dampers.
4. Dampers U and T will be blocked open and will no longer respond to closure signals.
During normal operation, filter trains M and N will be utilized together
()
with fans A, C, E,and F. During refueling operation, in addition to the equipment normally operating, filter train Q and fan G would also operate.
Am. 18
SUPPLEMENT 1, PART 2 RESPONSE TO QUESTION 52, PAGE 4 O
If contamination (radioactivity) is sensed in the ductwork, dampers S and V will close and Fin F will trip. This action will serve to separate the Auxiliary Building and Fuel llandling Building Ventilation systems and assure that all air leakage will be into the fuel handling building.
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t's' SUPPLE 3tE -- ),PART2 (v)
ATTAClBIENT 2 TO QUESTION 52
1. Fans: A,B,C,D,E - Existing F - Existing (Upgraded)
2. Filters: ht,N,0,P - Existing
3. Ducts: Solid Line - Existing Fuel 11. Blog. Co. trol aldg Aux. Bldg.
Broken Line - Existing (Upgraded) 4
4. Dampers: T and U - Upgraded l
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Air Fan Intake Jan M bt To Vent y -
5DDf Fans Filters AUXILIARY AND HiEI HANDLING BUILDING VENTILATION SYSTEh! PRIOR TO RESTART
g. _ _ _ _ __ .____ , .___ ._ _ - _ __ _ _ . _ _ _ _ . . - _._ ._ . . . _
SUPPLEMENT 1/"' ART 2 ATTACIMENT 3 TO dESTION52
1. Fans: A,B,C,D,E,F . Existing G,Il - New ESF Filter System Aux. Bldg. Fuel 11. Bldg. Control Bldg.
2. Filters: M,N,0.P - Existing Q,R -~New ESF Filter System
3. Ducts: Solid Line - Existing N Broken Line - New ESF
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4. Dampers: U and V - Existing T S and T - New ASF Filter r System new "lunnel" 1 V
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SUPPLEMENT 1, PART 2 O
\/v 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. See Section 7.2.4.
B. The permanent solidification system, currently being designed for TMI-1 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. Pt. waste) and 50 Cu. Ft. waste) per day respectively.
C. The PCP for the temporary systems being considered was forwarded to the NRC on May 21,1980 (GQL-TLL-237) . The design of the' pcrmanent system has not been finalized and a PCP cannot be drafted until the design is complete. The PCP will not bc finalized until the system is tested using non-radioactive test solutions and resin slurries.
For any system used to solidify radwaste, a PCP conforming to 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 temporary storage of EPICOR-1 prefilters and demineralizers. The combined EPICOR-I and EPICOR-II liner production rate will result in 6 to 7 storage cells being filled each month until the ~ -II auxiliary building water inventory is processed (approximately May 1980). After that time the production rate will be reduced to 3 to 4 filled storage cells per month. The current construction schedule for the completion of storage' cells will insure storage 7-( ,j capacity in excess of the production rate to the end of 1981.
Am. 18
SUPPLEMENT 1, PART 2 RESPONSE TO QUESTION 53 PAGE 2
%,1 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 configuration 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:
55 gallon drums unshielded (compacted trash) 200 maximum
a. l 50 cubic foot liners (evaporator bottoms unshielded) 8 maximum b.
or 50 cubic foot liners (used precoat unshielded) 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. .
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 cwft Compacted Trash (Dry) 1000 cwft Solidified Resin (Based on a normal operating year with refueling outage).
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Am. 18
SUPPLEMENT 1, PART 2 O
QUESTION
95. -
Paragraph 2.1.3.b of NUREG-0578 requires a description of. further measures and supporting analyses that will yield more. direct in-dication of low reactor coolant -level and inadequate core Section cooling2.1.1.6 such as reactor vessel water level instrumentation.
of the Restart Report does not address further measures (to be im-plemented by January 1,1981). nor does it address Provide the a conceptualquestion of reactor vessel water level Lastrumentation.
description of what additional measures will be taken to detect Provide an Duplementation schedule for inadequate core cooling.
these changes.
RESPONSE
Babcock & Wilcox has evaluated -the need for further measures beyond those described in Section 2.1.1.6 that will yield more direct A status indication reportof oflow their reactor coolant level and inadequate core cooling. We concur with review is attached (B&W letter dated April 4,1980, ESC-68).
B&W's report since there is no need for reactor vessel water level to determine inadequate core cooling. The basis for this statement is that reactor vessel I level is not used or neaded in order for the operatorIn to particular, take necessary there is corrective action to assure adequate core cooling.
no additional corrective action that can be taken that would not already t-() have been taken in response to a more direct measure of inadequate core cooling; namely, core
. temperature.
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(g) SUPPLEMENT 1, PART 2
ATTACHMENT TO OUESTION 95 Babcock &Wilcox %.w.mm. c,m C' O.O. Box 1260, Lynchburg. Va. 24505 Telepnone: (804) 384-5111 Apr il 4,1980 ESC-68 File 582-7102-T1.1
Subject:
Report on Additional Instr mentation to Detect Inadequate Core Cocling
Reference:
Letter, R. E. Davis (B&) to 3W Owners' Group
" Inadequate Core Coolin, Instrumentation,"
" ESC-11, January 15, 19E)
Gertlemen:
N lased for your review and commeit is a copy of the report docu-menting the investigations perforntd to satisfy the requirements of Section 2.1.3.6 of NUREG-0578. Th.s report concludes that additional instrumentation systems are not required to detect inadequate core ,
cooling and that there is no system available to meet the NRC cri- .
teria for supplementary instrumentation. It is recommended this re-p port be submitted to NRC following review.
U -
W have limited the scope of this report primarily to those potential-instrument additions based on principles which are commonly employed for reactor coolant system indications. Numerous other techniques have been suggested by various sources to meet the NRC criteria.
. Most of these are discussed in the "Pror eedings of the U. S. Nuclear l Regulatory Cocnission Review Group Conierence on Advanced Instru=ent'a- )
tion for Reactor Safety Research," NUREG-CP0007, October, 1979. Some ~ l of these techniques show pro =ise of providing a means of reactor ves- l sel water level indication, and our present understanding is that sev- I eral of the more promising techniques are being pursued under N?4 24D sponsorship. Unfortunately, based on the criteria that the indication i taist 1) hr tude th. full range from normal operation to below the core, 1
2) not erreneously indicate "-dequate cere cooling by unreinted phe-nomenon, and 3) provide an advance ward ng of the approach to inade-quate core cooling, none of the above :echniques can presently meet these requirements. Unless the crite .is is altered, a complex system using several technique.s coupled wit.a a signal analysis system, yet to be defined, would be required to meet the 'riteria.
We have reconsidered, therefore, the recocmendation ,>rovided by the reference letter to install hot leg level measurement capable of men- '
suring the delta-pressure between the bottom and the top of tae het
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leg. The difficulties involved in the installation of such a device and the difficulties involved in the interpretation of such a signal indicate that it may be more prudent to investigate more sophisticated systems if additional instrumentation or further synthesis of existing
'"'**-a in in&ed required. It is still our opinion that instrumenta-f tion to indicate the approach of inadequate core cooling would be use-ful in control of an accident situation. We have ceased work on the operating guidelines and hardware supply for the hot leg delta-pressure instrumentation under the current inadequate core cooling program. As you are aware, we are proceeding with work on operator guidelines for the use of high point RC system vents. Should this work indicate a need for bot leg delta-pressure measurement, or other additional instrumenta-tion, you will De informed under that program.
We are proceeding with development of a more detailed justificat. ion for the use of existing instrumentation in detecting and responding to in-adequate core cooling. This is based on the work performed to develop the additions to the small break operator guidelines and the analysis done for these guidelines and is intended to provide additional assur-ance that present readout capabilities are adequate for the required op-erator actions. This is in addicion to the criteria being developed for incore thermocouple aadouts, decay heat pump protection, and reactor vessel level measurement during refueling conditions. We do not intend O to further eursue additiena11nstrumentatien i= thia erostam =nt11 the requirements are better defined.
Please contact the writer if you have any questions on the above or
. on the enclosed report.
Sincerely, AM l -
Q William J. Keywor Product Manager .
WJK:jsp Attachment ,
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O STATUS REPORT ON ADDITIONAL INSTRUMENTATION TO DETECT
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-O + O TABLE OF CONTENTS O
2
1. diTRODUCTIGN
2. SUMfMRY OF RESULTS .
3. DISCUSSION 3.'1 Level Measurement concepts 3.2 Ultrasonics 3.3 Nuclear Radiation Methods 3.4 Additional Thermocouples .
3.5 Differential Pressure 4.
Q CONCLUSI0iLS
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1. INTRODFCTION The major concerns raised in the aftermath of the THI-2 accident are identified 4
Section in the "TMI-2 LESSONS LEARNED TASK FORCE STATUS REPORT, NUREG-0578".
2.1.3.b of that repcrt addressed the additional instrumentation which may assist
" i tinn em this; ariditional in the detection of in W .u?te cc~ tra' %
instrumentation was stated as:
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" Licensees shall provide a description of any additional instrumentation or controls (primary or backup) proposed for the plant to supplement those devices cited in the preceding section giving an unambiguous, easy-to-interpret indication of inadequate core cooling. A description of the functional design requirements for the system shall also be included. A description of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a m
schedule for installing the equipment shall be provided."
)
This position was clarWied and amplified in Enclosure 1 to H. R. Denton's i
' letter to all operating nuclear power plants of October 30, 1979 entitled I " Discussion of Lessons Learned Short Tem Requirements" as follows:
I "1. Design of new instrumentation should provide an unambiguous k This may require new indication of inadequate core cooling.
[
measurements to or a synthesis of existing measurements which meet safety grade criteria.
2. The evaluation is to include reactor water level indication.
3. A commitment to provide the necessary analysis and to study l
- advantages of various instruments to monitor water level and
$ core cooling is required in the response to the September 13, 1979 l
letter.
4. "The indication of inadequate core cooling must be unanbiguous, in fh I
that, it should have the following properties:
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' a. it must indicate the existence of inadequate core cooling caused by various' phenomena (i.e., high void h fraction pumped flow as well as stagnant boiloff).
b. it must not erroneously indicate inadequate core cooling because of tha crucare M = unraleted ph u w..ictiv .
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5. The indication must give advanced warning of the approach of .
inadequate corc cooling.
6. The indication must cover the full range from normal operation to complete core uncovering. For example, if water level is chosen as the unambiguous indication, then the range of the instrument l (or instruments) must cover the full range from normal water leve.1 to the bottom of the core."
2.
SUMMARY
OF RESULTS O
We have completed our investigations into the various means of detecting -
inadequate core cooling including' level measurement, and the following summarizes the results of our investigations:
, a. The investigation showed that the existing instrumentation will detect inadequate core cooling. The hot and cold leg RTD's along with pressure -
indication will indicate whether the RCS is subcooled or saturated. This determination'will lead to operator. actions designed to return the RCS to a subcooled condition. -Should these actions be ineffective, the incore thermocouples or the hot leg RTD's indicating superheated temperatures for the existing RCS pressure will provide a positive indication that the core is partially uncovered and inadequately cooled. Those indications i
lead to operator actions designed to reflood the core region. These' b operator actions will cover the core and begin to reestablish RCS
, inventory. The existing. instrumentation will'show when.the core is
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b. An evaluation.of a reactor coolant system water level measurement has also been completed. The result of this investigation is that a water
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level measurement that meets all of ina r.n t =n a 2n v n= e lsr 4 T-1 to the Lessons Learned Task Force Report cannot be developed'by the
- end of 1980 as required by NUREG-0578. The methods investigated eithe-
1) did not cover the entire range to the bottom of the core, or 2) contained uncertainties large enough to cause the measurement to be considered ambiguous or 3) involved techniques that required a more lengthy R&D effort than could be accommodated within the available time.
As a result of this investigation, and because the core exit thennoccuples can unambiguously, detect inadequate core cooling, no additional instru-mentation to measure reactor coolant system water level is being proposed A
U at this time. .>
3. DISCUSSIONS 3.1 The various methods of instrumenting a level measurement system that were considered are:
, a) Ultrasonics _
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b) Neutron or Gamma Beam i c) Additional Thermocouples J
d) Differential Pressure .
l- .The potential capabilities and problems with each measurement technique i are discussed below.
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, 3.2 Several methods of ultrasonic techniques have'been considered. They
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, include using existing internal structures as wave guides, installing e
an externally excited ultrasonic vibrating rod and installing a head mounted transducer. All of these methods have an excellent theoretical
- - . w 4 However, in the base and have been proven in simple applications.
reactor vessel, the core provides a heat source which changes the
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'" density of the fluid in the reae.or vessel.
The fluid may change state from a single phase liquid to a two phase fluid, and finally to a single phase vapor. Ultrasonic level measurement techniques are frequently A
f usea where tnere is a sharp density change at the fluid interface.
level created in a reactor vessel as a result of a LOCA will be a frothy two phase mixture height rather than a fixed level. 'In order to provide f -
an easy-to-interpret indication, a lengthy R&D effort would be required to insure that the output signal is unambiguous under all conditions.
In addition, all of the ultrasonic methods except using existing internal structures as a wave guice appear to require a new reactor vessel penetra-tion. Since neither the research nor the vessel penetration could be accomplished by the specified installation date, further investigation of ultrasonic techniques was terminated.
Q 3.3 Neutron and gamma beams have been used successfully to determine the level of fluid in a vessel. The application of this method to an RV level would be to use the core as a source and use the existing out of core detectors to monitor the water level through changes in count rate.
Normally, the detector count rate falls at rates characteristic of the various mechanisms of neutm production that exist following a reactor trip.
The source range detectors will respond to a decrease in water density through several mechanisms. The predominant effect is improved neutron However, if the water level transmission from the core to the detector.
drops below the top of the cbre, the other effects may dominate, causing '
a decrease in detector output. Resolution of the uncertainty in inter-l preting the behavior of the source range detectors would require an l
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extensive testing and calibration program which would exceed the speci" led installation date. For this reason, further investigation i
of this method was terminated. A more detailed discussion of the application of this nuclear radiation method is included in B&W document ~86-1105508-01.
3.4 Additional thermocouples installed axially in the incore instrument guide tubes were considered to provide an indication of the extent of the inadequate core cooling. This method is viable to give a direct indication of the portion of the core which is inadequately cooled.
iiowever, this instrumentation is not being proposed because it is not considered relevant information for the operator. An indication that the middle of the core is inadequately cooled will not elicit any further operator action over and above the actions taken when the top of the core indicates inadequate core cooling.
3.5 The use of differential pressure trbnsmitters to measure level was also considered. Three level measurement ranges, one across the reactor vessel, a second across the hot leg, and the other combining the two ranges, were evaluated.
A reactor vessel differential pressure (dp) measurement will require new penetrations in an incore nozzle at the bottom of the reactor vessel and at the top in a CRDM closure. An instrument can be insta1Ted to provide a differential pressure between the bottom of the core and the top of the l
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- reactor vessel, but the dp will be effected by not only the water level j head, but by shock loss, friction loss, and flow acceleration loss.
j During forced flow conditions, the shock loss, friction loss, and flow acceleration loss terms dominate the signal. Additionally, the magnitude j/
it of these terms varies depending on the density, and thus flowrate, of the
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. U V l pumped fluid. Due to the changing magnitude of those tenas, it is not possible to compensate them out of the dp signal to achieve a u) water level head only. During stagnant boiloff, the decay heat in the core will cause the level of coolant in the core region to swell to a level greater than that in the downcomer region of the reactor vessel.
A dp level neuurement will meoure Die culiopsed levei, Umi, in Ute downcomer region. A swelled level of 12 feet will still cover the top.
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of the core and remove heat. However, a swell level of 12 feet could indicate a collapsed level of between 7.4 and 8.625 feet depending on system pressure. The unpredictabic peak power distribution and decay heat level preclude compensating that level error out of the dp signal.
4 I9 Although the parameter of interest in this case is the mixture height, N the dp cell will measure a collapsed level which means that under certain conditions, this signal will be ambiguous.
A hot leg differential pressure measurement will require new penetrations j u> ,
(, in the bottom of the hot leg and in the vent line at the top of the hot I ~
leg. Again, this instrument would provide a dp signal and not an actual water level. But, in this instance, measuring any water level is a valid j
I indication that the core is covered. During flow conditions, the output signal would be affected by the same effects as the reactor vessel da t
f signal. However, the hot leg dp signal can be temperature compensated.
n The primary drawback to this measurement is that it does not meet the fv i requirement to measure level over the f ull range to the bottom of the core.
r; A differential pressure measurement from the bottom of the reactor vessel to the top of the hot leg will require new penetrations in an incore nozzle at the bottom of the reactor vessel and in the vent line at the top of the hot leg. This range is a combination of the two previous i
a instrument ranges. It pmvides an advantage over the hot leg level J
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. . O t' measurement in that the reactor vessel will not have to be defueled in order to install the lower tap, but it will still exhibit the same 7
In addition, O ambiguity as the reactor vessel dp described earlier.
due to the greatly expanded range, the inaccuracy of the instrument will be greater, perhaps as largn .s ' 1.0 ' t TH: re.mmnt n;;Ie :: r.:::e n n :.; r.: T.;'. I;; rr.;;. ;n . a. .,c a , Goo . .. U.
teactor vessel range.
4. CONCLUSIONS The existing instrumentation in the B&W designed plant is able to detect inadequate core cooling. The incore thermocouples are the primary indicators that inadequate core cooling has occurred. The hot leg RTD*s can be used as a secondary indication. These instruments provide an unambiguous indication of the existence of inadequate core cooling caused by various phenomena, but r~~, will not erroneously indicate inadequate core cooling. An advance warning of w;
the approach of inadequate core cooling is indicated by loss of subcooling.
The incore thermocouples can be used to indicate that the core is inadequately cooled regardless of the extent of core uncovery.
l Each system level measurement concept fails to meet at least one of the iestablished criteria for detection of inadequate core cooling. First, level measurement will not give a direct indication of inadequate core cooling.
fore cooling is indicated by temperature measurement, not level f.easurement.
I Secondly, all of the level measurement concepts, except additional thermocouples, d
pvide an output signal which may either be an ambiguous indication or will fail to provide indication over the entire range to the bottom of the core.
The information added by additional, axially placed, incore thermocouples makes I ,
mo difference in the management of an accident which has caused inadequate core mling and provides no advanced warning of a condition of inadequate core j tooling. Therefore, none of the additional instrumentation concepts which have i
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1 SYSTEM DESIGN DESCFJPTION FOR HIGE FRESSURE INJECTION CROSS-CONNECT p THREE MII.E ISI.AND UNIT #1 V
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Am. 18
1 TABLE OT CONTENTS . ,
Page_
~ section _
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1.0 Increduction 1 Summary 2 2.0 Discussion 3 3.0 Background 3 3.1 3.2 Calculacional Techniques 3 3.3 Systen Perfomance Under Cold Lag Break Condicions 3 3.4 Core Flood Line Break 4 f .
3.5 High Pressure Injection Line Break 4 i
3.6 Normal Plant Operations 5
3.7 Transienc System Operacions 5 3.8 Core Cooling Using Only High Pressure Injection 6 l
4.0 Operator Accion 7 5' O Post Modification Testing 3 j 5.1 2aquired Test Equipmenc 8 l
5.2 Procedure Abstract 8 5.3 Acceptance Criteria ?
6.0 References 10 9
9 Am. 18
= _ . _
l LIST OF FICUT.E5 -
Figure No. Title O .
1 Flow Diagram for Pump "A"
! 2 Tiow Diagram for Pu=p "C" 3 Simplified Systen Schema:ic of HPI Cross Connect 1
LIST OF TABLES Table No. Title 1 Eigh Pressure Injection Flow Requirements for Small Break LOCA's 2 liigh Pressure Injection Requirements for Core Flood Line Break ia h
' ~ ~ ~
3 Eigh Pressure Injection System Perfomance With Valves N!7-V16 f A & 3 Open and MU-P1A operacing 4 High Pressure Injection Syscam Perfomance With Valves MU-16 C & D Open and MD-Plc Operating v s 5 High Pressure Injection System Perfomance With Two Pumps and All Valves Open 6 High Pressure Injeccion System Performance Under High Prussure Injeccion Line Break Conditions e
o Am. 18 v .
I l , - _ . _ _ -.,
1.0 Introduccica To improve the ability of 7.II-l to withstand the consequences of a small break Loss-of-Coolant Accident (LOCA), a change in the design of the O
hi.s ,ress re inse tion 7ste- s de eto,ed. 2he seh;ec= chanie in o1 es 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 connece design.
As a result this report also represents the system description for the TMI-1 high pressure injection system.
O. .
i 6
4 e
O -
1- Am. 18 j
.. - . .- =. ._ . - - -. . . - _.
1 1
1 j 2.0 Summarv i The criterion established by B&W for the small break analysis requires
()
that- 70% of the :ocal flov for one RPI pump be injected into the unbroken legs of the reactor coolant system. .This criteria applies to a 2772 Mu ther'mai 177 fuel-assembly plant. For TMI-l with a licensed core power of 2535 Mut, 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 IP.I-1, the acceptable flow split can be relaxed
! to 64% - 36%. The above criteria are applicable to all breaks except for those reall breaks which occur as a result of a break in the high pressure t injection line between the RCS and the first isolation check valve in the
injection leg.
i
The analysis presented in this report demonstrate that the proposed
! cross connect at TMI-1 can meet the 3&W ECCS acceptance criteria.
. The analysis of this report further demonstrate that if the small break LOCA results from a break in the HPI line, sufficient high pressure injection flow will , occur through the unbroken EPI lines to satisfy small break ECCS criteria, i
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Am. 18
3.0 Discussien 1
3.1 Background _
A mechanical piping cross connect has been accepted as a viable long-term
( solution for additional corrective action on the small break I.CCA. The 1]. basio, concept is derived from the 3SAR 205 plants. ne 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 tfI-1 are detailed by CAI drawings C-302-661 Rev. IA-3 and E-304-666 Rev. IA-3. The following discussion presents One system flow calculations performed in support of the subject: design changes.
3.2 Calculational Technicues A simplified sketch of the proposed t1I-1 cross connect is shown in Figure 1 and Figure 2. Based on these layouts, a task was initiated to decemine 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 EPI line break. .
CAI's "PIFF" computer code was used to model the system. The PIPF Code is described in Topical Report GAI-TR-105Np-A. The subject report has been submitted to the NRC and accepted.
3.3 System Performance Under Cold Leg Bre:ak Conditions The initial performance of the til-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 VfI-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 tiI-I design, 3&W was requested to provide the HPI flow assumptions utilized in their small break analysis code. Reference (b),(o) and (r) transmitted this information.
- The criteria of reference (b),(o) and (r) are contained in Table 1 and are based on degraded pump performance. For the small break ECCS calculations, a factor of 0.9 was applied to subject flows during the first 10 ninutes of the transient to account for pump degradation due to wear. The resulting degraded pump performance is also presented in Table 1, Column B and forms the basis for Figure 6.2.59 of reference (c) .
As explained in reference (1), the operator action was assumed at 10 ninutes to increase the HPI flows to the intact cold legs. This was to be accom-plished by balancing one HPI pu=p to the four delivering lines, thus reducing the total effective resistance to the flow. Therefore, pump flow increased and pump degradation was implicitly taken into account. The resulting flows are presented in reference (o), and are those specified in Table 1, Column A.
, The above assumptions on flow have also been assumed in the analysis of ref. (r) .
'- B&W was also asked to specify the max 1=um pressure difference which would exist between the unbroken cold legs and the cold leg containing the bresk.
It was reported this value would be less than 4 psig. ,_
- ^** 16 Based on tha abovc, GAI has parformed calculations for cach of ths RC pressures specified in reference (b) . For conservatism, the shortcot 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 h,
pressure injection pump was assumed operating.
mance was assumed.
Undegraded pu=p perfor-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 run-out considera-tions. Run-out flow for the IMI-l high pressure inj ection pumps is con-sidered to be slightly greater than 550 gym 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 spm) when only a single pump is operating. Specifically, they limit flow to 137.5 gym when the venturi inlet pressure is 313.6 psia. At this flow, the vendor indicates that the non-recoverable pressure losses are 15% of the inlet pressure.
The GAI calculations, therefore, assumed that the n= * 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 3 and 4 presents the calculated results of the TMI-l high pressure Table 5 pre-injection system performance under the above assumptions.
sents the system performance when cth pumps are running. The acceptanco 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 MW plant and by the power ratio of TMI-l to the generic plant design (i.e. 2535/2772).
As indicated by Tables 3 and 4, the TMI-l cross-connect design with cavi-tating venturi not only meets the above acceptance criteria but also =eets the 70/30 flow split criteria for a 2772 Mwe plant.
g V' Core Flood Line Break _
3.4 The core flood line break establishes the maximu.a size acceptable for a cavitating venturi. Under these accident conditions, low RCS pressure occurs and high pressure injection is required. The cavitating venturi should, there-fore, limit flow to less than run-out flow of a singic pump. Run-out flow for the TMI-1 high pressure injection pumps is considered to be slightly greater 550 gpm since high flows have not yet been demonstrated by test. The flow assumed in the BFW Core Flood Line Break analysis is listed in Table 2 and as can be observed is less restrictive than that used in the Small Break analysis.
As indicated in Section 3.4, the venturi has been sized based on id'd ting flow to 137.5' gym when the inlet pressure is 813.6 psia. Based on this size venturi, enle"l ations 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 gym asseing the highest head pu=p is in operation.
3.5 High Pressure Inier:1on Line Break _
One of the small breaks considered in this evaluation wasSuch that associated a break if with the break of a high pressure injection line break.
it were to occur between the RCS cold leg and the first isolation check I
valve would result in both a small break LOCA and a EPI line break.
l 1
j fm Calculations performed on the original design (i.e. the design without C cavitating venturi) indicatad that the HPI line break required the operator Am. 18
~ ~~'
to isolcto tho Icg contcinits tho high ficw. Such action gecs egainst the cperccces comal ecadcccy end therofora pecscnts tha 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. ,
Rose devices will eliminate the need for the operator to take action which goes against his normal judgement.
Based on the venturi parameters presented in Section 3.3, the performance of the EPI system under high pressure line break conditions were analyted.
The results of the analysis are presented in Table 6. *hese results'were then transmitted to B&W. Their review and conclusions are documanced in reference (o) and reference (q), and indicate that the design performanca of the system will be acceptable. Table 6 and Figure 4 su:mnarize the ac-captance criteria contained in 3&W references for this break and demon-strates the acceptable performance predicted for the system.
3.6 Normal Plant Operations Due to space limitations at IMI-1, the high pressure injection line cross connect can only be accomplished inside of the reactor building. ,
ne cross connect will, therefore, be installed downstream of the locatien where under normal operations reactor coolant system =ake-up was being supplied. This resulted in an unacceptable system design since due to the cross connect, oscillating make-up flow between the "3" and "D" legs could occur and result in high cycle themal fatigue failure of the injection nozzles.
To correct this problem the normal make-up injection point to the "3" 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" EPI line downstream of the cross connect. A check O valve will be added to the "B" EPI line to prevent back flow of normal make-up water to the cross connect. The existing containment isolation valve, MU-V18, will provide 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, pressurizar level decreases anu operator action is takan. This action currently consists of opening MU-V16B and starting a second make-up pump as necessary to restore pressw izer level. This action is required because normal maka-up control v.1ve, MU-V17, significantly restricts the maximum deliverable flow. Ihe 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" EPI 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 noz=le without continuous flow). -
Q ^" 18 As a rocult, en altcrnste mecas for initicting high maka-up flow to the "B" HPI leg hco bocn provided by tho design. Spacifically, bypesa valva, MU-V217, is to be installed around MU-717. 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-V163 and allow the operator to deliser a maximum of 450 gpm of make-up.at 1800 psig RC pressure backpressure. Flow will only be through the "B" EPI nozzle and therefore thermal shocking will not occuri To provide indication of flow, a flow mater is being installed on the bypass line. The flow meter is a scrap-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.S Core' Cooling Using only High Pressure Injections i 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 feedvater is lost. To provide this back-up cooling capability, B&W analysis indicates that one HPI pump capable of injecting 216 gpa 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 safec'y factor of 1.0.
The TMI-1 cross-connect design has been analyzed to ensure that the cavitating venturi due not restrict flow below 216 gpm. The results of the CAI analysis indicate that the system will be capable of injecting a' minimum of 253.9 gym at an RCS pressure of 2500 psig. Therefore the
s- back-up machod of core cooling using the HPI system will be available.
r J
Am. 18
4.0 coarscor Acticn .
Following actuation of High Pressure Injection, no specific operator action should be necessary to achieve proper flow and flow split condi-fw tions. The only i= mediate action required by the operatdr is to verify V that at least the mini =um 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 A of Table 1. -
2. If both HPI pumps are operating, two or more HPI valves must have full open indication except that an indication that only valves MU-V16A and 16C or MU-V16B and 16D are 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-acion. In this event the operator must diagnosis the problem and take correceive action (i.e. start idle HPI pumps, open closed MU-V16 valves, etc.).
Ol e -
em S.
f e
8 6
4 0
/
Am. 18
5.0 Past Modification Toscing The following tests of the high preszure cross-connect design are recommended prior to the return of IMI-l to power operations.
5.1 Recuired Test Eouiement
1. Wide range RC Pressure - RC3A-PT3, RC3A-PT4 and RC33-PT3; range o to 2500 psig.
2. EPI Flow - MU23 - DFT1, 2, 3 & 4; range 0 to 500 gym.
3. Temporary HPI Injection Lag Flev Macers - TI - 1, 2, 3, 4 (See Figure 3) consisting of:
a. Flow Display Ccmputer: Controltron P/N 24IN - 2.5SS.375; Range O to 330 gym.
b. Multiplexer - Manual Selection 4 dhannels Controltron P/N 242-10.
I
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 PI2, MU22 PI3; range O to 5000 psig.
5. Make-up Pump Suction Pressure: PX-412, PX-413, PX-414 5.2 Procedure Abatract
1. Establish RC pressure at less than 600 psig and take adequate pre-cautions for overpressure protection (See Tech. Spec. 3.1-2 and Tech.
Spec. Change Request 74).
2. Ensure minimum recire. line is open s u,qi to be operated 'rui pressurizer level is being maintained between 90 .o 115 inches.
3. Start MU-P1A 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 pot i tion and establish maximum letdown.
NOTE The cavitating venturis were designed to prevent inadvertent pump run-out and cavitation without the need for throttling MU-V16 A through 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.
CAUTION Extended operations of IIPI 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 1200 ft. Record flow at.which unstable conditions are g reached. U 1
i Am. 18
CAUTION :
Do not operate any other make-up pump while minimum flow recire. ;
line is secured. Do not close MU-V16A and 3 while recire. line is !
. isolated and pump is running. !
5. Record flow through MU23-DPT16 2 and through each sonic flow indicator.
Record RCS pressure, pump suction and discharge pressure.
Observe venturis and cross-connects for signs of any unacceptabic vibration.
6. Reduce flow to less than 4d0 gpm and open MU-V36 and MU-V37. Close MU-V16A/B.
7. If necessary, align system so that RCP seal injection is being supplied from MU-PIC. Verify MU-V36, MU-V37 and MU-V64C are open. Start MU-PIC. Stop MU-PIA (B). ,
8. Repeat steps 3 and 4 except use valves MU-V16C and D and MU-PIC in lieu of MU-V16A/B and MU-P1 A(B) .
9. Record flow through MU23-DPT3 & 4 and through each of the sonic flov ,
meters. Record pump suction and discharge pressures.
10. Open or verify open MU-V36, MU-V37, and MU-764C.
t
11. Secure testing unici 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-PIB if not already started and open MU-V16A and MU-V16B.
O
13. Shut MU-V36 or MU-V37. Observe caution of step 4.
14. Record flow through MU23-DFT 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-V163.
5.3 Accepeance criteria Step 5 - Flow in all combinations of three HPI legs shall equal or exceed 70% of the flow in Table 1, Column A. Total flow of all four legs shall be greater than 500 gpm and less than 550 gpm. No unacceptable pipe vibration is allowed.
Step 9 - Flow in all combinations of three HPI legs shall equal or exceed 70*6 of the flow in Table 1, Column A. Total flow of all four 1 cgs shall be greater than 500 gpm and less than 550 gpm. No unacceptabic pipe vitration is allowed.
Step 14 - Flow in all. combinations of three HPI 1 cgs shall equal or -
exceed 70% of the flow in Table 1, Column A. In addition, 4 if total flow of all four legs is less than 95% of the predicted value based on Table 2, an engineering evaluation of the data shall be requested to verify proper system performance.
Am. 18
6.0 Refrrrn vs (c) B&W (G. T. Fairburn) latece TMI-79-17 daccd Fcbruary 5,1979 latter to R. M. Klingaman, RE: Three Mile Island Nuclear Station - Unic 1 Small 3reak Analysis.
(b)' B&W (G. T. Fairburn) letter IMI-79-74 dated May 21, 1979 to J. 7.
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. d. 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.
Potes. ,
(f) B&W 1etter dated May 1,1978 from J. T. Janis to W. E. Potts, Re:
TMI-l Small Break Analysia.
(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 CQL #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 Seoall Break LOCA.
(m) Met-Ed letter GQL #2031 dated December 21, 1978 from J. G. Herbein to R. W. Reid (NRC).
(n) Mec-Ed letter GQL #2072 dated December 29, 1978 from J. G. Herbein to R. W. Reid (NRC).
(o) B&W (G. T. Fairburn) letter IMI-79-208 - dated December 21, 1979 to D. G. Slear, RE: Accessement of HPI System to Mitigate Small LOCA's.
c Am. 18 4
-- _ _ =. -. -_ -- - _ _ _ _ _ - _ - _ _ _ _ _ _ _
i rg (p) Met-Ed/GPUSC Report in Response to NRC Staff, " Recommended Require-
, V ments for Restart of Three Mile Island Nuc' ear Station Unit 1",
dated September 7,1979 as supplemented and amended.
(q) B6W (G. T. Fairburn) Ictter TMI-80-036 dated February 25, 1980 to D. G. Sicar, Re: Draft Input for SDD-211A, Rev. 1. ;
(r) BGW (G. T. Fairburn letter TMI-80-76 dated April 16, 1980 to D. G. Slear, Re: HPI Flow During Small Break Transients.
i
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$ Am. 18
Flow Diagra= for Pump "A" .
Lag "A" "B" "D"
~
Leg "C" Le Le~ -
p ; ,
d
/6 /\ -
L21=301.3 ft.
L22=237,by ft. L23=196.19'ft.
L2h=120.87 ft.
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FI-38k ( F7-386 )( FE-385
)( FI-387 ) ,
L11*=139.98 ft. eq. ;. .. . ,1 !C-V16A W-V163 M L12*=148.71 ft. eq. -
d = 2.125 in. * - -
d= 2.125 in.
O j 14723 )
(
's/ e i {1023 FE1 FZ2 J
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>- , RCP 3eal Injection L1=93.6 ft. eq. r
, d1=3.62k in.
m-v74A .
LIA=121 98 ft. eq.
d1A=2.62h in. .
.. a gy,.VT3A ' T-N l
. IC-P1A ,.
2 g%_, .
. *czclusive of valve equivalent ionsth ,
Am. 18 -
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, Flow Diagram ror ?u=p .
"C" : ..
. . ,' n - -
Leg '!A" Leg "C" Leg "3" Leg "D" .*
, g . . . .. . . . ..
A A A .If '. U.
L21-353 75 re: t22-185 re.- L23-2h9.69 c . Lab 67 37 c:. . . l.'f 4
f FI-38k) FE-296
.FE-385} ( 7E-387)l( _
2 1. 233.6 ft. es.
e . 2.12, m.
..p y .
-nec. .-v16D Tr q g g. 91f. .
s.
W23 MU23
{
hN_) - FZ3 ITh -
1 m e- ,
\
1 !
L146.52 . ft. eq, ,
d1=3.624 1.n. .
m-v74c .
L1A=121 98 ft. eq, .
d1A=2.62k in. .
MU-V73C .73,q
., ..n e .1 MU-Plc. ,, ,-
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,. l
exc1usive of valve equivalent length Am. 18
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O TABLE 1 I) ilIGli PRESSURE INJECTION FLOW REQUIREMENTS FOR SMALL BREAK LOCA'S ,
\
i j
Column A( ) Column B( )
Pump Flow Pump Flow RC Pressure After 10 Minutes First 10 Minutes (psig) (gpm) (gpm) 0 548 515 600 500 450 1200 438 380 1500 404 342 1600 391.7 328 1800 364.3 300 2400 260 210 2500 216 191.7 (I)0nly for small break conditions other than llPI line break. See reference (o) for flow assumed under llPI line break conditions.
(2)0nly 701, of this value was assumed delivered to the RCS.
( )0nly 501, of this value was assumed delivered to the RCS.
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+
Am. 18
._- - - . _ ... - . _ _ _ _ - - - _ - ~- -- -
O TABLE 2 ilIGil PRESSURE INJECTIGN REQUIREMENTS i
i FOR CORE FLOOD LINE BRE/K i
i
)
Flove to RCS
' RC Pressure (gpm)
(psia) 0 454.5 615 440 4
l
1315 365 l
I 340 g 1515 1615 l
i l
l i
i l
i
!O Am. 18
- . - -- - - . . - ~ - . _ - . . _ . . - - -- . - ..
<O O (o TABLE 3 llPI System Performance with Valves HU-V16 A & B Open and HU-PI A Operating RC Pressure in Psig IIPI Flow in gpm y ou in Unbroken Legsugpoi Calculated Seal Calculated Requi r ed( 1.) Total Flow ,
Leg "A" Leg "B" Leg "C" Leg "D" Hinimum (64%) gpse Imap "A" leap "B" Loop "C" Loop "D" Injection gpm (Three Leg Tot al) 0 0 127.8 127.0 126.8 126.0 34.3 381.6 352.1 541.9 0 0 539.2 300 500 496 130.3 130.6 128.5 128.8 21.0 389.4 500 600 596 125.8 131.9 129.1 130.7 18.3 386.8 320.1 535.8 600 600 1196 110.7 114.8 113.6 121.0 16.3 339.1 279.8 476.4 1200 1200 1200 1496 101.9 105.6 104.5 11i.8 15.1 312.0 258.8 438.9 1500 1500 1500 ,
1596 98.5 102.2 101.2 108.3 14.7 301.9 250.6 424.9 1600 1600 1600 1
1796 91.5 94.8 93.9 100.9 13.7 280.2 233.2 394.8 1800 1800 1800 2396 64 .9 67.1 66.8 73.4 10.2 198.9 166.4 282.4 2400 2400 2400 I
2496 59.1 60.9 60.7 67.3 9.5 180.7 138.3 257.5 2500 2500 2500 N
E (1) Refer to Section 2.0 for basis.
7-O O O
~
TABLE 4 HPI System Performance with Valves HU-V16 C & D Open and HU-Plc Operating s
RC Pressure in Psig HPI Flow in gpm Flow in Unbroken Legs, l e Cal ("lat'd Seal Calculated Requ,reJ(1), Total Flow.
Losp "A" Loop "B" 1Aop "C" Loop "D" Leg "A" Leg "B" Leg "C" Leg "D" Injection Hinimum (64%) spa EPs O' O O O 132.8 132.9 132.8 132.7 N/A 398.5 352.1 531.2 600 600 600 596 132.8 132.9 132.8 132.7 N/A 398.5 320.1 531.2 5200. 1200 1200 1196 111.4 114.6 119.4 126.6 N/A 34 5.4 279.8 472
.1500 1500 1500 1496 102.5 105.4 109.9 116.9 N/A 317.8 258.8 434.7 1600 1600 1600 1596 99.2 102.0 106.4 113.2 N/A 30 7.6 250.6 420.8 1800 1800 1800 179e 92.1 94.6 98.7 105.5 N/A 285.4 233.2 390.9 2400' 2400 - 2400 2396 65.4 66.8 70.3 76.7 N/A 202.5 166.4 279.2 2500 2500 2500 2500 59.9 61.9 64.5 67.6 N/A 186.3 138.3 253.9 U
(1) Refer to Section 2.0 for basis.
._ - -_ . - - . - ..- - .~. . . _ . - . . - _ _ . .. - - . . . . . - - . - . . . - -- - - .- -_. - - .. ~.
(O O 'O TABLE 5 IIPI System Performance With Two Pumps Operating and All valves Open RC Pressure in Paig HP1 Flow in gpm Flow In Unbroken lei;segga n Calculated Seal Calculated Requ. red (l) Total Flov "A" "B" Leg "C" Leg "D" Minimum (64%) gym 142p "A" Loop "B" Loop "C" Loop "D" Leg Leg Injection '
EPs t 0 0 199.1 199.0 199.l. 199.0 51.8 597.1 352.1 848 0 0 600 600 600 199.4 199.7 199.4 199.7 42.3 598.5 320.1 840.5
600 1200 1200 1200 201.2 201.0 200.9 201.7 29.8 603.1 279.8 834.6 1200
~
188.4 196.3 197.4 204.9 23.2 582.1 258.8 810.2
'l500 1500 1500 1500 1600 1600 1600 181.9 189.3 190.5 200.3 22.5 561.7 250.7 784.5 1600 1800 1800 169 175.9 177.1 186.2 20.9 522 233.2 729.1
, 1800 1800 2400 2400 2400 119.7 124.7 125.6 132.1 15.2 370 166.4 517.3 2400 l 2500 2500 2500 108.6 113.2 114.0 120.1 13.9 335.8 138.2 469.8 2500 N
E (l) unfer to section 2.0 for basis.
_ _ - - _ _ . _ - . - _ _ _ -_. --- _ _~ _
gG LJ (Q)
(f]
v TABLE 6 -
IIPI System Performance Under itPI Line Break Conditions s
Opercting RC Pressure in Psig IIPI Flow in gpm Flow in Unbroken Lega Calculated Nakrup ,
Calculated Required I , 2) Total Flow Fump Loop "A" Loop "8" Loop "C" Loop "D" Leg "A" Leg "B" Leg "C" les "D" Hinimum (64%)(3) spo C 0 0 0 0 132.8 132.9 132.8 132.7 398.5 320/336.5 531.2 C 600 600 600 0 132.8 132.9 132.8 132.7 398.5 268.6/305 531.2 C 1200 1200 1200 0 103.1 89.8 110.5 172.0 303.4 195.4/264.7 475.4 C 1500 1500 1500 0 90.1 64.4 96.6 188.9 251.1 154.6/N/A 440.0 C 1600 1600 1600 0 85.4 54.4 91.6 195.3 231.4 137.2/N/A 426.7 C 1800 1800 1800 0 76.2 33.0 81.9 206.4 191.1 102.9/N/A 397.5 C 2400 2400 2400 0 28.4 0.0 30.7 233.9 59.I N/A/N/A 293.0 (1) Refer to Section 2.0 for basis of 64% flow split.
(2) First value is required flow during first 20 minutes. Second value is required flow af ter 20 minutes.
(3) Acceptance criteria are documented in reference (o).
=,
Y
g ATTACINENT TO SUPPLEMENT 1, PART 3 QUESTIONS 1, 2 6 3 Babcock &Wilcox ro.e, ceneret,on c,

ep n
V P.O. Box 1260, Lyncheurg, Va. 245o5 Telephone: (804) 384. 5111 4 .
April 16, 1980 TMI-80-076 Mr. D. G. Slear (3)
TMI-I Project Engineering Manager GPU Service Corporation 100 Interpace Parkway 1 Parsippany, NJ 07054
Subject:
HPI Flows During Small Break Transients
Dear Mr. Slear:
Attached for your information and use is the information requested by your Mr. J. F. Fritzen and the i1RC during a telecon on March 4,1980. Attachment A documents the itPI flows used by B&W in performing the core flood line break analysis. Attachment B provides a writeup concerning the applicability of the I:PI flows used by B&W in the small break analysis relative to the TMI Unit I HPI flows. Attachment C provides suggested corrections to Table 1, Column B of System Design Description for High Pressure Injection Cross-Connect, 500 211 A,,
Rev. 1.
If you have any questions or require additional information, please advise.
Very truly yours, d
, v g
lsMr tW' G. T. Fairburn Service Manager GIF/cw cc: J. G. Herbein L. L. Lawyer J. J. Colitz J. F. Fritzen T. J. Toole P.. W. K e aten - GP' JSC F. R. Faist J. C. Lewis - Phila. Sales O l h B R:xk E.V/Aox Compe.ny / Estabbshed 1857
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ATTACIDIENT A 4
ItPI Flows During CFT Line Break i l 8 . l i Flow, r,pm Pressure, psia i
l 454.5 0.0 440. 615.
l 365. 1315.
340. 1515.
i i 325. 1615.
325. 3000. ,
1 l
4 1
l t
) .
I
!S -
1 1
l 1
}
l l
l I
i 1
l a b
(
i
.i O
9 i
y-.m v ,y -
m,-. . _ _ , , - -_ . _ - ,, ._ _.-m ,,--..,3.,_., w_,,.c _,,w. wor- - - -
--m-m-+ www.v---,e %w-re-+ +- - - + --e- e ,. -w-e-ww_v e - e* e
1 ATTA_CmtENT B Os_/
Applicability of the Sno11 Break B&W Model to the TMI-l Plant Characteristics Oa lbrch 6,1980, a joint telephone conversation took place with CPU, NRC, and B&W, concerning the TMI-l HP1 system. The NRC raised the question of the applicability of the B&W small break model (references 1, 2, and 3) to TMI-1, since the flows delivered by the T}R-1 HPI system are lower than those assumed in the B&W analyses for pressures below 600 spig (reference 2).
The HPI flows used by ESW are listed in the attached Table A. Flous in Column I are used for the first 10 minutes of the transient. Following manual cross-connecting of the HPI valves at this time, the flows in Column II are assumed for the remainder of the transient. The difference arises because, for a pressure of 0 psig,the TNT-1 system delivers 500 gpm instead of deliver-ing a flow equal to or greater than the 548 gpm analysis value. This is shown in Table A.
\
[v~)
It should be pointed out that the spectrum analyses docunented-in references 2 and 3 were performed at 2772 MW:, while the rated power of the "MI-1 plant is 2535 MWt. Table B lists the HPI flows which should be used following operator action at 600 seconds, if the 1MI-1 initial core power were nodeled in order to calculate the same system responses presented in reference 2. It can be seen that at 0 paig, the HPI flow should bc 501 gpm. Therefore, there is a discrepancy of enly 1 gpm, between the injection flow needed at 0 psig, and the TMI-l HPI flous delivered at that pressure.
In order to assess the impact of the smaller HPI flous in the analyses-de-scribed in reference 2, two of the cases presented there have been re-evalu-ated using hand calculaticas. Specifically, the 0.15 ft2 break at pump dis-charge has been considered since this break depressurizes the system to 600 ;
2 psig more quickly. The second break to be studied was the 0.07 ft at pump ;
discharge, which turned out to be the worst case. l l
(~)
s-l
1 2
Results of the 0.15 ft at pump discharge small break analysis show that the n prinary system pressure reaches 600 psig at about 500 seconds, 1.c., before the operator action (Figure 3, reference 2). At this time the IIPI pump is delivering 450 gpm (Table A, Column I), of which 50% enters the system, and the other 50% is lost through'the break. This pressure also coincides with the amtuation of 2 core flood tanks, which, f or this case, inject an average of ,about 55 lbm/s of additional flow to the reactor vessel. Core boil-off is matched in the analysis at about 550 seconds.
The mixture height ja the reactor vessel at this time is about 1 ft above the top of the core. Following operator action at 600 seconds, which results in an increase of the flow delivered to the reactor vessel, the mixture level in the vecsel starts to increase. Assuming, for conservatism, the small break model power of 2772 Init, as opposed to 2535 Init, the mixture level in the vessel is estimated to increase at a slightly slower rate than calculated in the analysis, due to the difference in HPI flows between the model and the TMI-l plant.
The core never uncovers, thus preventing any cladding tempera-ture excursion, assuring conformance uith 10 CFR 50.46.
O V
The 0.07 ft2 break at PL, itch, in reference 2 was found to be the worst cmall break, resulted in a minor core uncovery and a peak cladding tempera-ture of 1095F, which is well below the 2200F criteria of 10 CFR 50.46. Figure 3 of reference 2 shows that the system reaches 600 psig at about 1500 seconds, which also corresponds to the minimum core mixture Icvel (reference, Figure 4).
The top of the core is re-covered at about 1800 seconds. Using TNI-l flows, and the codel core initial power of 2772 Snit, a deficiency of about 350 pounds of liquid water exi:tr in the vessel at this time, a: compared to liquid in-ventory calculated in the analysis. Because of this difference, the vessel refills to the top of the core with a 15 seconds delay. Thus the peak tem; erature remains in the neighborhood of 1100F, as calculated in reference 2, ensuring compliance to the 10 CFR 50.46 criteria.
Reference 3, Appendix C, documents a core flood tank line break (0.44 f t2) analysis, performed for a typical 177 low-loop plant, having an initial power
.of 2772 FCit. The llPI flows assumed in that noalysis are lower than the FPI flows used in reference 2 (see Attachment A). /"ain, the calculated nystem L._. _
4 i :,
l . .
i ,
responscr. following, the initiation of the transient comply wholly with the r l
I g 10 CFR'50.46 criteria.
i 1
In sumary, the TMI-1 IIPI system, when used in a single failure mode in an hypothisized small break LOCA, can safely mitigate the transient within the limitations of Appendix K.
i e
I b l t
i i
I f
i .
l l i t i
l lo l
i i
l i
I n
O l O
i
- ;i
), .e ,
k Table A. IIPI F1nti 2 A'intind in the IIPI Stra11 Break Model l , Assumed llPI flows 1 Assumed IIPI flows the first 10 min after 10 min '
~
i (50%'is lost (30% is lost ,
Pressure, psig through the break) through the break)_ .
- 0 515 548 .
i 600 450 500 438
~
1200 380 -
i 1500 342 -
404 1600 328 391.7 1800 300 ,
364.3 2400 210 260.0 j . .
l 1 Table B i O -
~~
. Assumed IIPI fiou for TMI-l -
af ter 10 min (30% is lost Pressure, psig through the break) 0 501 600 457 1000 400 .
1500 369 i
2000 358' 2400 333 -
! REFERENCES k
i 1. Topical Report BAU-10104, Rev. 3, "UsW's ECCS Evaluation Model," August 1977."
Letter f ron J.it.- Ta) lor : (G1U) to S. A. _Varga -(NRC), July IR.1973.
2.
3. Topical Re;) ort' E.W-10103A, Rev. 3, "ECCS Analysis _of B&W's 177-FA
..l.owered-1.oop NSS," July 1977..
5
_A_TT6Cgg][T C, (g"%
. Tabic 1. IIPI Flow Requirements Column A Column B Pump flow Pump flow RC pressure, psig undegraded, Epm degraded, gpm 0 550 515 600 500 450 1200 437.1 380 1500 404.3 342.
1600 391.4 328 1800 364.3 300 2400 260 210 2500 216.0 191.7 0
l
( }Only for small break conditions other than a HPI line break. See reference (0) for flow assumed under HPI {
I line break conditions.
l
SUPPLEMENT 1, PART 3 rS b
QUESTION
11. The long-term requirement of IE Bulletin 79-05C requires the B6W licensees to submit a design which will assure automatic tripping of the operating reactor coolant pumps (RCPs) under all circumstances in which this action may be needed. It has been shown through analysis that this trip is needed for a certain spectrum of small break LOCAs.
Prior to final design acceptability, the following conditions must be satisifed:
a. Characteristic curves for RCP current / power versus void franction fraction must be fully demonstrated and documented based upon existing test data and supplemented as necessary with confirmatory data obtained from future tests such as LOFT, full scale testing, etc.;
b. Justification for the RCP current / power setpoint must be ,
shown; and
c. Satisfactory responses to the following must be received.
RESPONSE
(_/
The compicxity and schedule of responding to items (a) and (b) above and the desire to complete this modification in a timely manner have led us to reconsider the selection of RCP power as a trip parameter. Accordingly, we have determined that RCP trip on ilPI (Safety Feature Actuation Signal) with coincident loss of subcooling margin is an appropriate alternative to our previous proposal. Our previous proposal had called for a RCP trip on llPI with coincident RCP current / power setpoint actuation. With regard to design criteria for the revised RCP trip design, the following will be utilized:
Design Characteristics
1. The system will meet the requirements of IEEE-279.
2. Environmental qualifications will b' based on post-LOCA and post-steam line break accident environments.
3. Trip logic shall be single failure proof assuming the worst case single failure for all reactor coolant pump operating combinations.
4. No single failure will cause the spurious tripping of the reactor j coolant pumps.
v i
l Am. 18 I I
i
. SUPPLEMENT 1, PART 3 .
1 l l
1
- PAGE 2 TO QUESTION 11 f
Operational Characteristics 4
I
1. The setpoint of the monitor will be equivalent to zero fahrenheit
degrees margin, plus allowance for instrument error coincident with a 1600 psig SFAS signal.
2. Response time of the system will be testable from the outside of the
[ RTD thermowell through closure of the output contacts. . Response
, time will be no greater than the response time of the existing instrument channels.
i 3. Accuracy of the complete instrument train shall be + 5'F.
4. Bypass capability should be provided when pressure is below 1650 psig to allow restart of the pumps during startup. The bypass will be
automatically removed above this pressure. Note

Depending on j system design, this requirement may be met by the existing 1600 psig SFAS bypass.
! 5. Override capability should be provided to allow pump operation during inadequate core cooling conditions.
! The major advantage of the revised design is that the use of measured process
variables (reactor coolant system temperature and pressure) together with an accurate correlation to determine saturation conditions, and associated
, subcooling margin, provides for conceptual design simplicity. The use of RCP power / current correlated to void fraction is more diftleuit due to the s uncertainty associated with local voiding conditions.
i f
i f
4 Am.-18 y -y v + - ,--y ,- em-,-. , , --.E...- -- #T -*' - ' ' tW-rT 1 7 w-
SUPPLEMENT 1, PART 3 4
4 O CUESTION
12. By letters dated August 31, 1979, each B&W operating clant licessee indicated a general endorsement of S&W's generic report BAW-1564,
" Integrated Control System Reliability Analysis."
i
! Our joint review of this report witn Oak Rioge National Laboratory has progressed sufficiently to assure ourselves that tne recom-mencations that the report offers with regard to potential areas of improvement in ICS reliability are reasonable. Tnerefore, we request
, that you address these recommendations and discuss your followup action plans in this matter. Responses to the following items must Me provided.
/ As part of the continuing review of this report, additional areas
'_ / may be hignlighted as requiring improvement. In that event, we will provide additional requests in these specific areas as necessary.
RESPONSE TO 12.a.1 See Supplement 1 Part 2 response to Question 38.
RESPONSE TO 12.a.2 C To be submitted at a later date (September 30, 1980)
RESPONSE TO 12.a.3(a)
In the past, feedwater oscillations have occurred during two modes of operation:
4
1. Transition from startup to main feedwater flow control.
2. At power levels less than full power (60 - 75%) due to coupling with thg heater drain system.
Tne problems during feerdwater control transition occur due to leakage through the main feedwater control valve while the valve is closed (n'ot abnomal for i a modulating valve). As the startup valve reaches a predetemined percentage open, the main block valve opens to permit transition to main valve control.
Leakage results in excessive flow causing the startup valve to close, resulting in reclosure of the main block valve. This problem has been corrected by rechecking startup valve limit switches each refueling outage.
Also, some coerators make the transition with the startup valves under manual control, which is acceptable.
I i Oscillations due to coupling with the heater drain system have been minimi:ed by system tuning. This tuning has eliminated oscillations at full power.
However, changes in system dynamics at reduced powc? result in some oscillations in the range of 60 - 75% power. These oscillations are not a problem during
O-Am. 18
12.a.3(a). Succlement 1. Part 3 Continued power reductions cue to the short time that the plant is keot in :ne (s,) affectec range. During startup with power holds in the range of 60 -
75%, it may ce necessary to place reactor controls in Hanc to prevent un-necessary cycling of control red drives. Other actions taken to reduce oscillations have included installation of hydraulic snuocers on some heater drain system control valves, and tuning of the level controllers on the heaters and on the 6th stage drain collection tank.
Otner recur-ing actions taken to reduce feedwater system croblems incluce precycling of main, startug and block valves prior to plant startup to prevent sticking, and overnaul of one main and one startup feecwater valve each refueling outage.
RESPONSE TO 12.a.3(b)
The general operating philosophy for the ICS is to maintain all control stations in the automatic mode during steady state and transient operation.
The operator may intervene wnenever he judges that system operation is abnormal, or is inadequate to prevent exceeding reactor trip setpoints.
RESPONSE TO 12.a.3(c)
Procedures used by the operator to perform the operation described above are as follows:
1. Plant emergency procedures address symptoms of abnormal system
O behavior and . instruct the operator to intervene to acnieve the desired result.
2. Operating procedures for specific evolutions, such as pl: .: ::artup and shutdown, address manipulation of ICS controls. These procedures allow operator intervention if conditions appear abnormal.
3. The ICS operating procedure gives details of how and wnen to manitulate controls. Specific guidance is given for operating subsystems under manual control. , Again, operator intervention is permitted if conditions appear abnormal.
RESPONSE TO M.a.3(d)
Training covering the operation of the Integrated Control System was covered during the '" Operator Accelerated Retraining Program (OARP)". Additional training covering tne new procedure guidelines for the coeration of the ICS in hand (Procedure 1105-4, Appendix II) has been scheduled.
RESPONSE TO 12.b.1 l
To be submitted at a later date (June 30,1980)
RESPONSE TO 12.b.2 wO To be submitted at a later date (June 30,1980) !
RESPONSE TO 12.b.3 To be submitted at a later date (June 30, 1980)
Am. 18
/
o l
i l
I SUPPLEMENT '
l l
1 I
OPERATIONAL QUALITY ASSURANCE PLAN '
O FOR 1
THREE MILE ISLAND NUCLEAR STATION )
I UNIT 1 j l
l O l l !
.. , _ . - _ _ - _ _ _ _ _ _ _ _ . _ __ . . _ _ _ _ _ _ . . _ __

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