ML19347B807
| ML19347B807 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 10/10/1980 |
| From: | Pollard R UNION OF CONCERNED SCIENTISTS |
| To: | |
| Shared Package | |
| ML19347B808 | List: |
| References | |
| NUDOCS 8010160013 | |
| Download: ML19347B807 (22) | |
Text
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6-. O fS'*D 0 UNITED STATES OF AMERICA 3 kc, /, F NUCLEAR REGULATORY COMMISSION g : ['es,,'$}
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In the Matter of )
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METROPOLITAN EDISON ) Docket No. 50-289 COMPANY, -et _al., )
(Three Mile Island )
Nuclear Station, Unit )
No. 1) )
)
DIRECT TESTIMONY OF ROBERT D. POLLARD ON BEHALF OF THE UNION OF CONCERNED SCIENTISTS REGARDING UCS CONTENTION NO. 5 l l
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ROBERT D. POLLARD QUALIFICATIONS ,
Mr. Pollard is presently employed as a nuclear safety expert with the Union of Concerned Scientis ts , a non-profit coalition of scientists, engineers and other orofessionals supported by over 80,000 public sponsors.
Mr. Pollard's formal education in nuclear design began ;
in May, 1959, when he was selected to serve as an electronics technician in the nuclear power program of the U.S. Navy. .
After completing the required trair.ing , he became an instruc-tor responsible for teaching naval personnel both the theore-tical and practical aspects of operation, maintenance and repair for nuclear propulsion plants. From February, 1964 to ,
April, 1965, he served as senior reactor operator, supervis-ing the reactor control division of the U.S.S. Sargo, a nuclear-powered submarine.
After ..is honorable discharge in 1965, Mr. Pollard attended Syracuse University, where he received the degree of Bachelor of Science magna cu m laude in Electrical Engi-neering in June, 1969.
In July, 1969, Mr. Pollard was hired by the Atomic Energy Commission (AEC), and continued as a technical exoert with the AEC and its successor the United States Nuclear Regulatory Commission (NRC) until February, 1976. After joining the AEC, he studied advanced electrical and nuclear engineering at the Graduate School of the University of New Mexico in Albuquerque. He subsequently advanced to the oositions of Reactor Engineer (Ins trumenta tion ) and Project Manager with AEC/NPC.
As a Reactor E ng in eer , Mr. Pollard was primarily respon-sible for performing detailed technical reviews analyzing and evaluating the adequacy of the design of reactor protec-tion systems, control systems and emergency electrical power systems in proposed nuclear facilities. In September 1974, he was promoted to the position of Project Manager and became responsible for planning and coordinating all aspects of the design and safety reviews of applications for licenses to construct and operate several commercial nuclear power plants. He served as Project Manager for the review of a number of nuclear power plants including: Indian Point, Unit 3, comanche Peak, Uni ts 1 and 2, and Catawba, Units
! 1 and 2. While with NRC , Mr. Pollard also served on the i
standards group, participating in developing standards and j safety guides, and as a member of IEEE Committees.
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t i OUTLINE - DIRECT TESTIMONY l ON UCS CONTENTION NO. S i
This testimony discusses six specific safety-related functions of the pressurizer PORV, block valve and associated circuits and controls: 1) functioning as a portion of the reactor coolant pressure boundary, 2) limiting challenge.s to the safety valves, 3) preventing overpressurization of the reactor coolant system and, thus, damage to the reactor vessel during low temperature operations, 4) reducing challenges to ECCS, 5 ). " bleeding" cooling water during the bleed and feed mode, 6) depressurizing the reactor coolant system during 4
conditions of inadequate core cooling. It demonstrates that 1 the PORV, block valve and circuitry are important to safety i
l and, therefore, must be required to meet all applicable safety-1 j grade criteria.
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! UCS CONTENTION NO. 5 Proper operation of power operated relief valves, associated block valves and the instruments and controls for these valves ;
is essential to mitigate the consequences of accidents. In addition, their failure can cause or aggravate a LOCA. Therefore, these valves must be classified as components important to safety and required to meet all safety-grade design criteria.
Introduction The opening of the power operated relief valve (PORV) and its failure to reclose were key factors in the TMI-2
. accident. In addition, for several hours the operator failed to detect the open PORV and terminate the loss of coolant accident by closing the block valve. As a result of these events, the Commission ordered certain improvements or upgrading of the PORV, the block valve and the instrumentation and controls for these valves. I conclude that, considering the 1;ssons learned from the TMI-2 accident, the ordered improve-ments are necessary, but not sufficient to provide adequate 1
protection for the public. I conclude that the PORV, block 1 valve, and associated instrumentation and controls must be l 1
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5-2 classified as components important to safety and must meet all safety grade design criteria in order to provide reasonable assurance that operation of TMI-1 will not pose an undue risk to public health and safety. I also conclude that, in view of the events that occurred during the TMI-2 accident and the lessons learned from the accident, proper application of Commission policy and practice requires this safety grade
, classification.
[ To explain the bases for these conclusions, I will address each function performed by the PORV and block valve. Within each such discussion, I will address the relevant TMI-2 lessons learned and the position held by Met Ed and the Staff.
l However, first I will describe the purpose and operation of the PORV, block valve and safety valve.
. The PORV and safety valve are connected to the top of the pressurizer. If reactor coolant system pressure increases to their respective set points, the PORV and safety valve I
are supposed to open, thereby releasing steam hnd/or water from the reactor coolant system and limiting further pressure increase. When reactor coolant system pressure decreases below their respective set points, the PORV and safety valve are supposed to reclose.
The PORV is electrically controlled. An electrical signal derived from a measurement of reactor coolant system t
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- pressure commands the PORV to open and close, but does not
! directly open and close it. In contrast, the safety valve l is opened by reactor coolant system pressure acting directly i
- on the valve. The safety valve is designed to reciose after j reactor coolant system pressure decreases below the opening pressure set point.
l The pressure at which the PORV is commanded to open can be easily adjusted because it is electrically controlled.
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In normal operation, the PORV is automatically controlled by i
l a pressure instrument and is commanded to open when reactor i coolant system pressure increases to 2450 psig. The operator can also command the PORV to open at any pressure by manually controlling the electrical signal to it. In contrast, the i
safety valve cannot be controlled at all by the operator nor 1
can its opening pressure set poinu be changed during plant l operation. It is set to open at 2500 psig. The operator
! can neither open nor close the safety valve.
There is a history of relief and safety ' valve' failures
- in operating plants. The failures experienced have included opening below the set point, not opening at the set point, and not reclosing after pressure has decreased below the opening l
set. point. (See NUREG-0578, p. A-7) . The PORV has associated I
i with it a block valve which is located between the PORV and the pressurizer. By closing the block valve, the operator can 4
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i j 5-4 prevent the PORV frcm opening or can stop the loss of coolant j through the PORV if it fails to reclose. The block valve is l
manually controlled by the operator. The safety valve has l
no block valve associated with it and, therefore, the operator cannot terminate the loss of coolant through the safety valve if it fails to reclose.
i Safety Related Functions of the PORV The following are the primary safety-related functions of the pressurizer PORV:
- 1. The PORV is part of the reactor coolant pressure boundary. Therefore, an inadvertent opening of the PORV constitutes a loss of integrity of the pressure boundary which amounts to a LOCA or aggravates an existing LOCA.
- 2. The PORV serves the purpose of limiting the number of times the safety valve is called upon to open. Both relief and safety valves have an alarming history of failing to reclose. The PORV is equipped with a block valve so that it can be isolated if it fails to reclose. l
- 3. The PORV is used to prevent overpressurization of the reactor coolant system during low temperature operation i
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(e.g. start-up and shut-down) when the nil ductility transition i
temperature of the reactor vessel becomes the Ibniting consider-j ation for maintaining the structural integrity of the vessel.
- 4. The PORV block valve serves to reduce the challenge rate to the emergency core cooling system (ECCS), because i
inadvertent opening of the PORV or the inability to isolate the PORV requires ECCS to function.
- 5. The PORV is used to " bleed" cooling water during the licensee's so-called " bleed and feed" mode.
- 6. The PORV is essential to depressurize the reactor i
coolant syated in order to utilize the low pressure injection system during conditions of inadequate core cooling.
I will now discuss each of the above in some detail.
- 1. Reactor Coolant Pressure Boundary - ,
The Staff, Met Ed and UCS are all in agreement that the PORV is part of the pressure boundary and that a stuck-open PORV causes or aggravates a LOCA. However, the Staff and Met Ed propose to classify neither the PORV or its associated
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- circuitry as safety-grade.
The Staff has previously acknowledged that the probability of failure of the PORV in the open position " contribute (s]
significantly to the probability of a small break LOCA.
(NUREG-05 65, p. 3-7 ) . Thus, the relatively high probability of PORV failure is a significant contributor " the risk of a LOCA. Met Ed has taken no exception to this observation.
The defense-in-depth principle, a regulatory cornerstone of both the AEC and now the NRC, requires both that the probability of a LOCA be kept low and also that adequate protection to mitigate a LOCA be provided. This reflects the recognition that it is not desirable to place a reactor into conditions requiring use of emergency systems. Repeated challenges to emergency systems are unacceptable. Both aspects of this defense in depth policy are reflected in the General Design Criteria (GDC), Appendix A, 10 CFR Part 50. For example, GDC 14 requires that "[t]he reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to A single failure in the PORV circuitry could cause the PORV to open inadvertently. I have noted that, although NUREG-0578 (S2.1.2) specifically calls for qualification of the control circuitry associated with the PORV, the TMI Restart Evaluation, (p. C8-10) does not include this requirement.
" Licensee's Response to UCS Letter Dated August 19, 1980,"
September 2, 1980, p.4.
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l have an extremely low probability of abnormal leakage, of l ,
l rapidly propogating failure, and of gross rupture" (emphasis l
j added), while other GDC such as 35, 36 and 37 set forth the l
l requirements for ECCS. The existence of ECCS does not in any way relax the extremely strict requirements for integrity of the reactor coolant pressure boundary.
In contrast, Met Ed and the Staff appear to take the position that so long as there is a system (ECCS) designed to cope with a LOCA, it is not necessary to prevent the LOCA from occurring. If this position were to prevail, it could presumably also be argued that the reactor coolant pressure boundary need not meet GDC 1 (quality assurance), GDC 2 l (seismic qualification), or GDC 14, previously quoted. Indeed, I
the valve performance testing program and the direct indication i
of valve position ordered as short-term requirements by the l
l Commission would be unnecessary if Met Ed and Staff's l
arguments were accepted. The consequences of valve failures l can presumably be mitigated by ECCS.
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See NUREG-0578, Table B-1, items 2.1.2 and 2.1.3.a. These, like all of the Table B-1 items, were specifically adopted in the Commission's order of August 9, 1979.
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5-8 With respect to the need to meet the requirements applicable to equipment important to safety, there is no technical basis for the distinction which Met Ed makes between components which form part of the reactor coolant pressure boundary and other equipment important to safety. Components and equipment which form part of the reactor coolant pressure boundary are important to safety.
This can be clearly seen by contrasting the Staff 's stated position on the PORV with its position on the newly-required reactor coolant system high point vents, the purpose of which is to provide a mechanism for venting noncondensible gases.
The following is a quotation from page C8-60 of the TMI Restart Evaluation:
" POSITION Each applicant and licensee shall install reactor coolant system and reactor vessel head high point vents remotely operated from the control room.
Since these vents form a part of the reactor coolant pressure boundary, the design of the vents shall conform to the requirements of Appendix A to 10 CFR Part 50, General Design Criteria. In particular, these vents shall be safety grade, and shall satisfy the single failure criterion and the requirements of IEEE-279 in order to ensure a low probability of inadvertent actuation." (Emphasis added).
Using precisely the same logic, which is the correct approach, the PORV and its instrumentation and controls should be safety grade, and should satisfy the single failure criterion i
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" Licensee's Testimony of James H. Correa, Gary T. Urquhart l and Robert C. Jones, Jr. in Response to UCS Contentions 5 and 6", p. 7 and 8.
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f l and the requirements of IEEE-279.
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i 2. Limiting Challenges to the Safetv Valves i
One of the major design functions of the PORV was to I provide a means for controlling pressure in the primary system l without the need for reactor SCRAM - in other words, to enhance the " availability" of the plant in view of the sensitivity of the B&W design to operational occurrences which result in pressure transients. This was accomplished by having i
j the opening pressure set point of the PORV lower than the i
high pressure set point for reactor SCRAM.
- One of the lessons learned from the TMI-2 accident was that it is necessary to reduce the frequency of pressure l transients which would open the PORV or the safety valve.
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! IE Bulletin 79-05B required, inter alia, modifications to l reduce the likelihood of automatic actuation of the PORV
) during anticipated transients. It was also stipulated in j the Bulletin that the modifications not result in increased frequency of pressurizer safety valve actuation for these
! transients. This reflects a basic recognition of the inherent i
unreliability (or inadequate qualification) of the valves i shown through a history of valve failure.
The response to the issues raised by IE Bulletin 79- 05B
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were as follows:
- a. The high pressure reactor SCRAM set point was reduced to 2300 psig and the PORV opening set point raised to 2450 psig.
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- b. New reactor SCRAM signals will be installed to shut down the reactor in the event of turbine trip or loss l
1 of feedwater.
In my view, these modifications are not sufficient for the following reasons:
- a. The control circuitry for the PORV has not been classified as safety grade and does not meet the requirements of IEEE 279. In order to provide reasonable :
assurance that the FORV rather than the safety valves will open, the circuitry should be required to meet IEEE 279.
- b. Reducing the difference between the PORV and safety valve set points from 145 psi to only 50 psi makes it even more important that the PO,RV and its controls be of the highest reliability (i.e. safety-grade) in order to achieve the stated safety objective of not increasing the instances of safety valve actuation.
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- 3. Pressure Control During Low Temperature Operation During periods of low temperature operation such as
> l l start-up, shut-down, and recovery from. accidents, it is necessary l l .
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i to limit the maximum pressure in the reactor coolant system to a significantly lower value than the maximum pressure permissible at normal operating temperatures. At low temp-cratures, the steel of the reactor vessel is susceptible to cracking (i.e., brittle fracture). Until the reactor vessel walls are above the nil ductility transition temperature, the reactor coolant system pressure must be limited to a few hundred pounds per square inch.
The PORV is used during low temperature operations to protect against overpressuring the reactor vessel. This l function cannot be performed by the safety valves because i
the opening pressure set point - 2500 psig - is far above the permissible pressure and cannot be changed by the operator.
The ECCS are not designed to cope with a rupture of the reactor vessel, an accident the Commission has deered " incredible."
Thus, during low temperature operation, the PORV clearly performs a safety function. -
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- 4. Reducing Challenges to ECCS One of the important general lessons learned from the TMI-2 accident is the necessity of decreasing the frequency of occurrence of conditions or accidents which require the operation of ECCS. This is the topic of dicc2ssion at page j 1
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5-12 6 of NUREG-0578, for example. The goal of reducing ECCS challenges is important to safety.
Inadvertent opening of the PORV or the inability to isolate a stuck-open PORV by closing the block valve results in a plant condition requiring operation.of ECCS. As discussed above, neither the PORV, its block valve, nor its associated instrumentation and controls are safety grade. A single failure in the circuitry could result in inadvertent and inappropriate opening of the PORV. Therefore, to fully accomplish the safety function of reducing challenges to the ECCS, the valve and its associated controls and instrumentation should be safety-grade and meet the requirements of IEEE 279.
Similarly, because of the history of PORV's failing to reclose, the function of isolating a stuck-open PORV is also a safety function needed to achieve the safety objective of reducing ECCS challenges. Therefore, the block valve should i
be classified as a safety-grade component and its controls should comply with the requirements of IEEE 279.
In summary, the function of decreasing challenges to ECCS is a safety function and non-safety equipment cannot be relied upon to perform this function. Two pieces of non-safety equipment -- the PORV and block valve -- do not substitute for-or compensate for.the lack of a safety-grade component i
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and compliance with the GDC and IEEE 279. ,
lh I 5. Bleed and Feed Throughout its testimony, Met Ed relies on the so-called 4
"biced and feed" mode of cooling as its final line of defense.
- For example, the licensee states that, in the event of a i failure of feedwater for a small LOCA, heat removal can be
- accomplished by bleed and feed - a " form of forced circulation cooling." similarly, Met Ed claims that there is no need for modifying the residual heat removal system (low pressure injection) so that it is capable of functioning at reactor l design pressure because of the availability of bleed and feed.
Bleed and feed is raid to work in the following way:
4 the high pressure injection pumps are used to feed water into j the reactor coolant system and the water is bled frcm the system through the PORV into (eventually) the containment sump. This use of the PORV to perform the bleed functic'n is For purposes of this testimony, it will be assumed that bleed and feed is an effective method of cooling the core.
It is my opinion that this is not a valid assumption.
- " Licensee's Testimony of Robert W. Keaten and Robert C.
Jones in Response to UCS' Contention Nos. 1 and 2," p.7.
Id, p. 10-11.
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. specified in TMI-l Emergency Procedure 1202-6B, " Loss of j Coolant / Reactor Coolant Pressure (Small Break LOCA) Causing Automatic High Pressure Injection." A copy is attached.
NUREG-0578, page A-1, states that "[t]his method of de zy heat removal requires the use of the emergency core cooling system (ECCS) and the power-operated relief valves (PORVs) i j in the pressurizer."
Despite the fact that Met Ed intends to use the PORV during bleed and feed, and the TMI-1 emergency procedures
, _ for bleed and feed instruct the operator to use the PORV I
for this purpose, Met Ed now attempts to downplay the importance of this function of the PORV by stressing instead that the bleeding function of bleed and feed may be acccmplished by j use of the safety valves only, which are said to be safety grade.
While it may be true that the safety valves can be relied on during bleed and feed, their use has,significant disadvantages compared to use of the PORV. As noted earlier, i
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the operator can control the PORV whereas he has no control over the opening and closing of the safety valves. Furthermore, if in the course of the repeated opening and closing of the valve required for bleed and feed, it should stick open (by no means an improbable occurrence) , the operator can 4
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5-15 j isolate a stuck-open PORV and continue to operate in bleed and feed by using the block valve. In contrast, if the
, safety valve sticks open, the operator can take no action to regain control of the bleed and feed mode.
One of the major lessons learned from the TMI-2 accident I is that reliance on ..an-safety grade components to provide protection for public health and safety is insufficient.
Thus, for example, the auxiliary feedwater systems are now required to be safety grade. However, with regard to the
'l PORV, the Staff appears content to identify the problem, but not solve it prior to TMI-l restart.
"Several nonsafety systems were used at various times in the mitigation of the accident in ways not considered in the safety analysis; for example, long-term maintenance of core flow and cooling with the steam generators and the reactor coolant pumps. The present classification system does not adequately recognize either of these kinds of effects that non-safety systems can have on the safety of the plant.
Thus, requirements for nonsafety systems may be needed to reduce the frequency of occurrence of events that initiate or adversely affect transients "and accidents, and other requirements may be needed to improve the current capability for use of nonsafety systems during transient or accident situations." (NUREG-057 8, p. 18, emphasis added).
"The'relatively high frequency of AOOs places a reliability demand on the operation of the PORVs and associated equipment that is higher than originally envisioned.
Also, the operation of some components and systems provided for emergency core cooling have been challenged more times than was previously expected as a result of i
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5-16 AOOs. Therefore, there is a need to consider the upgrading of the PORVs, block valves, and the associated
! control and power equipment to a safety-grade classification l to achieve greater valve reliability and to minimize the number of challenges to the operation of the emergency core cooling components and systems. However, l
the merits and degree of upgrading of all pressure-l relief equipment associated with the pressurizer requires l further evaluation, which should be accomplished on a longer term basis." (NUREG-0578, p. A-3 and A-4) .
The bleed and feed cooling mode has been devised with i
reliance on the.PORV and all applicable procedures and operator training have been directed toward use of the PORV. Met Ed and the Staff apparently take the position that they may rely on non-safety equipment. They justify this by arguing that other, safety grade, equipment is available, but hopefully will not be needed. This is analogous to their previous position which resulted in licensing of the TMI plants with non-safety grade auxiliary feedwater systems. In my opinion, Met Ed and the Staff's position that the PORV need not be safety grade is contrary to the lesson to be learned from the TMI-2 accident, unnecessary, and has significant safety disadvantages. The
" bleeding" function of the PORV in the bleed and feed mode is i
l a safety function and the components should be safety grade.
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- 6. Deoressurization of Reactor Coolant System Under Conditions of Inadecuate Core Cooling
, Finally, there is a portion of the " inadequate core cooling"
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5-17 protection which relies solely on the PORV. It is described in Attachment 3, " Actions for Inadequate Core Cooling," to Emergency Procedure 1202-6B. This procedure instructs the operator to "[o] pen the pressurizer PORV and leave open."
The procedure notes that "[t]he RCS will depressurize and the LPI system should restore core cooling."
This function cannot be performed by the safety valves because the operator has no control over these; they will not open below 2500 psig and consequently cannot be used to depressurize the system. Other potential methods of depressuri-zation, such as use of the letdown line, would not be available under inadequate core cooling conditions (such as those existing during the TMI-2 accident) because of the high level of radio-activity in the reactor coolant system after core damage. This is explicitly recognized in Met Ed's emergency procedures.
(See Emergency Procedure No. 1202-39, " Inadequate Core Cooling,"
Step 39.3.2.g.) Thus, use of the PORV to depressurize the system under inadequate core cooling conditions is a safety function for which no alternative using safety grade equipment is available.
Emergency Procedure 1202-39 relies on the turbine bypass l valves, atmospheric dump valves, and the pressurizer heaters to achieve cold shutdown conditions when letdown cannot be used.
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5-18 Conclusion I have discussed above a number of functions of the PORV which are important to safety. Taken as a whole, these f present a clear and consistent picture: the PORV, block j valve, and the associated instrumentation and controls are safety-related and must be required to meet all app'icable i
safety-grade criteria before TMI-1 is permitted to resume q operation. Reliance on non-safety equipment to perform safety functions presents undue risk to public health and safety.
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f Attachment to " Direct Testimony of Robert D. Pollard on behalf of the Union of Concerned Scientists regarding UCS Contention No. 5."
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- FO 3 USE IN UNF FFUWW 1202-68 .
Revision 3 06/30/80 THREE MILE ISLAND NUCLEAR STATION '
UNIT #I EMERGENCY PROCEDURE 1202-6B -
LOSS OF REACTOR COOLANT / REACTOR COOLANT PRESSURE.
(SMALL BREAX LOCA) CAUSING AUTOMATIC HIGH PRESSURE INJE g g Taele of Effective Pages U$11N UNITI ONLY -
Paoe Date_ Revision Pace Date Revision _
Paoe Date Revision 3 51.0 1.0 06/30/80 3 26.0 06/30/80 52.0 06/30/80 3 27.0 06/30/80 , 3 '
2.0 3 53.0 3.0 06/30/80 3 28.0 06/30/80 54.0 29.0 06/30/80 3 4.0 06/30/80 3 3 55.0 5.0 06/30/80 .3 30.0 06/30/80 56.0 6.0 06/30/80 3 31.0 57.0 7.0 06/30/80 3 32.0 58.0 33.0 8.0 06/30/80 3 59.0 9.0 06/30/80 3 34.0 60.0 ;
10.0 06/30/80 3 35.0 61.0 11.0 06/30/80 3 36.0 .
62.0 12.0 06/30/80 3 37.0 63.0
-06/30/80 38.0 64.0 13.0 3 39.0 14.0 06/30/80 3 65.0 06/30/80 40.0 66.0 2
15.0 3 41.0
- 16.0 ~ 06/30/80 3 67.0 06/30/80 42.0 17.0 3 43.0 68.0 18.0 06/30/80 3 69.0 06/30/80 44.0 19.0 3 45.0 70.0 i
20.0 06/30/80 3 71.0 06/30/80 46.0
, 21.0 3 47.0 72.0 22.0 06/30/80 3 73.0 23.0 06/30/80 3 48.0 74.0 l 49.0 24.0 06/30/80 3 75.0 25.0 06/30/80 3 50.0 proval Unit 1 Staff Recor,m d Approval Unit 2 Staff Recom ends
- Approval Date Approval Date Cognizant Dept. Head Cognizant Dept. Head Unit 2 PORC Reco me s pproval Unit 1 POR Recommen s Approva! -
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- oat, (- 2 5- Po Chairman of PORC V- Chairman of PORC Unit 2 Superinterident i Unit 1 Su intendent A proval g // Date
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Manager Generatien Quality Assurance Approvgl M Date TMt 65 A Rev 8/77 60101600 L O R U SE I N U N IT I b \, ' _Y
,", fun UOL 11 M UTVT ITu rvto-
- 1202-6B
- Revision 3 06/30/80 THREE MILE ISLAND NUCLEAR STATION UNIT 1 EMERGENCY PROCEDURE 1202-6B :
LOSS OF REACTOR COOLANT / REACTOR COOLANT PRESSURE (SMALL BREAK LOCA) CAUSING AUTOMATIC HIGH PRESSURE INJECTION NOTE: (1) If RC pressure remains above ESAS setpoint proceed per 1202-6A.
(2) If OTSG tube rupture is confimed proceed per 1202-5B.
(3) If loss of coolant is large enough to initiate HPI, CF and LPI flow automatically and RC pressure drops below 200 psig, proceed per 1202-6C.
(4) The block diagram in Attachment 5 may be used as an aid in detemining the applicable procedure and section. The block diagram does not supersede any procedure steps.
6.8.1 Symptoms i
- 1. Decrease in reactor coolant pressure. RC Pressure Hi/Lo !
Alarm. l
- 2. Reactor Trip Alam.
- 3. Turbine Trip Alam.
- 6. Ap indication on the PORV or code safety valve discharge line or alarm on the acoustic monitor on the PORV.
1.0 FOR USE IN U \ T I ON LY
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,. 1202-6B ;
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l 6.B.2 Immediate Action A. Automatic Action
- 1. Reactor trip on icw RCS pressure or pressure to temperature.
- 2. Reactor trip containment isolation occurs.
- 3. ESAS actuation at 1600 psig.
B. Manual Action l NOTE: The parameters indicated with an asterisk (*)
will be reverified as the first step in Follow-Up Action.
- 1. Press the reactor trip push button and verify that all rods have inserted (except Group 8).
- 2. If RCS pressure decreases to HPI setpoint verify HPI automatically initiates, HPI pumps start, MU-V14A and B open and MU-16A, B, C and D open and trip all operating Reactor Coolant Pumps innediately following verification of HPI.
- 3. Verify the Turbine is tripped and verify that the Turbine stop valves are closed.
- 4. Close MU-V12. .
- 5. Close RC-V2 (may be reopened if necessary to control a pressure increase).
- 6. Verify that EF-P-1, EF-P2A and EF-P-28 start and discharge pressure increases to >1010 psig, EF-V'30A and B open and flow is indicated to increase steam generator level to 50%.
- 7. Evaluate for a Non-LOCA overcooling event by observing the folicwiiig:
FOR USE N2 U NIT I ONILY
. N \EfeAb UV\J N N 1202-6B .
0*o"/NO3 NOTE: During a loss of coolant accident, RCS conditions will approach saturation. -
Temperature will remain fairly constant -
and pressure will decrease to saturation pressure. During a Non-LOCA overcooling ,
event, pressure and temperature will both decrease rapidly. Some subcooling will be maintained unless pressurizer level is lost.
- a. RCS pressure, temperature, and saturation margin.
- b. OTSG pressure: Decrease for steam line break.
. c. OTSG 1evel: >50% for overfeeding event, low level for steam line break.
- d. Main and startup feedwater flow: High flow for overfeeding event, isolated for steam line break.
If non-LOCA overcooling is indicated proceed per Attachment-2.
- 8. Take manual control of EF-30A and B and slowly raise OTSG 1evel to 95% on operate range.
l
- 9. Verify that the reactor trip building isolation system has acutated.
- 10. Evaluate for OTSG tube rupture by observing the following:
3.0 l
l FOR USE IN U NIT I O\'LY l... . -
_ 9...... ...
U' ' " ' ' ' '
.~ FUN VOC llN .
. . . 1202-6B Revision 3 06/30/80
- a. Main steam line high radiation ~ monitors.
- b. Condensor air ejector high radiation monitor (RMA-5)
- c. Verify no increase in RC drain tank and reactor building temperature, pressure and radiation level.
If a tube rupture is confirmed, proceed per EP1202-5B.
- 11. Monitor reactor coolant drain tank level, pressure, and temperature for any unexpected increase. Check PORV and code safety valve discharge flow alarms in the control room, and check acoustic monitoring for PORV.
6.B.3 Follow-Up Action The objectives of the Small Break LOCA Emergency Procedure are to trip the reactor, determine whether a LOCA or a NON-LOCA overcooling event has occurred, conserve RCS inventory, keep 0
the core covered and cooled by maintaining 50 F subcooled condition, establish natural circulation and remove core heat
- through the steam generators, achieve a cold shutdown condition and to minimize radioactive release to the environment.
- 1. Reverify the parameters in Manual Action that are marked with an asterisk (*). Use redundant indication if available.
- 2. Announce on the page that a loss of coolant accident in Unit 1 has occurred.
l ,3 . Verify the following reactor building isolation valves closed.
FO R USE I N" U N F I O N LY
- ..a .
,: l-U M U d t- I W U INI I i VW'i
- - . 1202-6B Revision 3 06/30/80 NOTE: These valves automatically close on reactor trip.
WDL-V534 WDL-V535 t RC Drain Tank j WDG-V3 WDL-V303 WDG-V4 WDL-V304 l RCS Sample i i
CA-V1 CA-V3 I CA-V2 CA-V13 RB Purge AH-V1A AH-V1C AH-V1B AH-V1D Core Flood Tank CF-V2A CF-V2B CF-V19A CF-V19B CF-V20A CF-V20B Demin Water CA-V189 OTSG Sample CA-V4A CA-VSA CA-V4B CA-V5B Let Down Cooler MU-V2A MU-V2B
- 4. Verify that both trains of HPI components are operating per table 1. If both HPI trains have not actuated, attempt to start second HPI train.
FOR USE I N1'G N IT I O NLY ..
. . .f,.0... - -- -
- FOR USE IN UNII I UWL T 1202-6B '
Revision 3 .
l 06/30/80 l
J CAUTION: Do not attempt to operate ES valves to the Non ES position and do not trip any ES components
~
until the ES signal is bypassed. The breaker ,
i anti-pump feature will prevent reclosure of
)
tripped breakers until the ES start signal is I cleared.
- 5. As time permits connence plots of RCS temperature and i i
pressure as follows:
- a. On Figure 1 plot hot leg RTD temperature vs pressure.
- 6. If reactor building pressure reaches 4 psig, verify that the RB isolation valves have closed as indicated on Panel PCR.
RC Pressure, psig Required HPI flow gpm 0 550 600 500 1200 438 1500 405 1600 -
392 1800 365 2400 260 2500 216 l
FOR USE N"JNIT I ONLY l I
I-U H U d t- I N U nli i i vin i 1
- 1202-6B Revision 3 06/30/80 -
1
~
If flow requirements are not met verify valve positions and attempt to start another Make-Up Pt..np.
- 8. If inadequate core cooling is indicated by one or more of the following conditions refer to Attachment 3: 3
- a. Superheated conditions as indicated by the saturation meter.
- b. Superheated conditions as indicated by the RC hot leg RTD's and the narrow or wide range RC pressure indicators,
- c. Superheated conditions as indicated by the incore themocouples for the existing RCS pressure measured by the narrow or wide range pressure indicators.
- 9. Close or verify closed RC-V3, RC-V1, RC-V35, MU-V1 A, MU- f V1B and RC-V28 in an attempt to isolate the leak.
- 10. Evaluate radiation levels, implement the reouired actions -
and make appropriate notification in acccrdance with the Emergency Plan AP-1004, and event review and reporting requirements AP-1044.
- 11. Intemediate Closed Coolin; (IC) and seal injection should be maintained to the RC Pump seals to assure long tem availability. If NUC Service closed cooling to the RC Pump Motors is lost, monitor bearing and stator temperatures. Each pump should not be run for more than ,
30 minutes without cooling water to the motor.
- 12. When the 50 F subcooled margin is established, the RCS pressure /downcomer temperature shall be kept in the 7.0 i
FOR USE IN UNIT I ONLY
, F UWMN ;
. - 1202-6B Revision 3 06/30/80 Acceptable Region of Figure 2. HPI ficw may be throttled to achieve an acceptable pressure / temperature combination while maintaining a 50 F subcooled margin. Full HPI flow shall be reinitiated if the 50 F subcooled margin cannot be maintained.
CAUTION: Do not throttle HPI unless one of the following three conditions exists:
- a. The LPI system is in operation and flowing at a rate in excess of 1000 gpm in each line and the situatien has been stable for 20 minutes.
U
- b. All hot and cold leg temperatures are at least 50 F below the saturation temperature for the existing RCS pressure, and the action is necessary to prevent the indicated pressurizer level from going off-scale high. If 50 F subcooling cannot be maintained, full HPI shall be reactivated.
- c. Or, all indicated hot and cold leg temperatures are at least 50 0F below the saturation temperature for the indicated RCS pressure and continued full HPI injection will result in RCS pressure /downcomer temperatures within the Restricted Region of Figure
- 2. With no operating RC pumps the RCS downcomer temperature shall be detemined by subtracting 150 F from the average of the five lowest temperatures from operable incore themocouples. Incore thermo-couples with temperature indications 500 F below the ,
bulk T hot may be considered inoperable.
FOR USE lAoUNIT I ONLY
@ UDE TlWUNIh 1202-6B Revision 3 06/30/80 1
- 13. Verify that Natural Circulation has been established by -
l observing: .
- a. Cold leg temperatures approach saturation temperature for secondary side pressure within approximately 5 t minutes.
- b. Primary delta T(Thot - Tcold) becomes constant at approximately 300F.
- 14. Monitor Source Range Nuclear instrumentation and incore thermocouple temperatures. Unexpected increase in source range indication or incore thermocouple temperatures may indicate voiding. If voiding occurs assure that full High Pressure Injection is in operation.
- 15. Send an operator to the emergency feed pump area to take manual control of the EF Valves if problems occur. If neither main nor emergency feedwater is available, maintain full high pressure injection until feedwater is restored.
Open the PORV and RC-V2 or allow the code safety valves to open to provide a flow path. Refer to Attachment 4 for restoring Emergency Feedwater. -
- a. Defeat 2 channels of RB Isolation in Actuation A and !
B (required only if RB Isolation has occurred).
- b. Bypass all three channels for 1600 psi in Actuation A and B (If the RCS has repressurized to greater than 1700 psi reset the 1600 psi trip bistables rather than bypassing).
- FOR USE lY U NIT I ONLY .
P
FOR USE IN UNII
~
I VINL I 1202-6B Revision 3 06/30/80
- c. Throttle MU-V16A(B)(C) or (D) as needed to prevent the pressurizer from going solid. Monitor total -
make-up pump flow to maintain > 80 gpm per pump. I Open MU-V36 and MU-V37 if needed to provide bypass J
~-
flow.
CAUTION: Because MU-V12 is closed, monitor make-up tank level and pressure to avoid lifting the makeuo tank relief valve (MU-RV-1). _
- 17. Reduce turbine header pressure setpoint or take manual control of the bypass valves to cooldown the RCS so that Th is > 50 F below RCS saturation temperature per Figure i il and cooldown rate is approximately 100 0F/ hour but 4 within Tech Spec limits. When-50 0F subcooling is achieved l and aC pressure is below 700 psig close CF-VlA and CF-VlB to isolate core flood tanks.
- 18. If the criteria for RC pump restart per attactment 1 is l
met and if RC pumps and emergency feedwater are available, bump the RC pumps in accordance with Attachment 1, and proceed to forced cooling.
I 19. If RC pumps are not available continue cooldown on natural circulation.
i 6.B.4 Long Term ;
- 1. Upon a Borated Water Storage Tank Lo-Lo alarm of 36",
1 shift suction of MU pumps from the Borated Water Storage Tank to the Reactor Building Sump by opening DH-V7A and DH-V7B, DH-V6A and 6B and closing DH-VSA and SB and closing MU-V14A and 14B. Injection flow path is now as follows:
FO R USE IN)oleJNIT l ONLY
, P U M U C C llN vlNii i v'
- 1202-6B Revision 3 06/30/80 Spilled coolant to R.B. sump, R.B. sump to LPI pumps, LPI .
pumps to MU pumps, MU pump to RCS via MU-V16A, B, C, D.
- 2. When both MU pumps are in piggy back mode of operation with >500 gpm flow per pump and RC pressure is <250 psig trip all RC pumps (if operating). Monitor Hot Leg RCS temperatures and incore thermocouple temperatures. l
- 3. If RCS pressure decreases to <200 psig, and the conditions of the caution following Step 6.B.3.12 have been met, i throttle HPI discharge flow by throttling MU-V16A/B/C/D.
If required open MU-V36 and MU-V37 to provide at least 80 i i
gpm thru each HPI pump. Periodically open MU-V12 as necessary to control makeup tar.k level and pressure.
i CAUTION: Because MU-V12 is closed, monitor make-up tank
)
level and pressure to avoid lifting the makeup )
l tank relief valve (MU-RV-1). l l
Observe that LPI pumps deliver water to RCS via DH-V4A/B
(>1000 gpm per LPI loop). l
- 4. When LPI is verified >1000 gpm and stable-for >20 minutes in both loops, stop the make-up pumps and close DH-V7A, 7B and MU-V16A/B/C/D. Injection flow path is now as follows:
Spill coolant to RB sump, RB sump to LPI pumps, LPI pumps to RCS through DH-V4A and B.
- 5. Throttle DH-V4A and 4B as required to maintain LPI pump flow of >1000 <3500 gpm and when time permits, throttle l
11.0 FO R USE IN U NIT I ON LY
t U M U KfWUW FFFUI\'JU al
~
-1202-6B Revision 3 06/30/80
~
manual valves DH-Vl9A and 19B and reopen DH-V4A and 4B.
Within about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> establish a long tenn cooling circulation mode as described in OP 1104-4 and listed below.
Mode 1 - Forced circulation using decay heat drop line.
Mode 2 - Gravity draining reactor coolant hot leg to the Reactor Building sump via the D.H. drop line.
Mode 3 - Hot leg injaction using pressurizer auxiliary spray line.
Mode 4 - Reverse flow through the decay heat drop line into "B" reactor coolant loop hot leg.
- 6. Monitor reactor building hydrogen concentration. Place the hydrogen recombiner in operation if hydrogen concentration reaches 3% per OP 1104-62.
- 7. Containment isolation valves may be opened to obtain samples in accordance with approved procedures. The isolation valves shall be reclosed after the sample is obtained.
- 8. Other containmen- isolation valves automatically closed shall remain closed until the follcwing conditions are met.
- a. Reactor building pressure is less than 2 psig.
- b. Containment radiation levels have been assessed based on radiation monitor readings or samples.
l l
l 12.0 FO9 USE lh UNIT I ON LY
[ -
FWWETWUNFFFUNL T
~
1202-6B
~
Revision 3 06/30/80 ;
J
- c. The integrity of the system outside the reactor building has been assessed. (Stable surge tank .
level, visual inspection or pressure test should be considered to verify integrity). $
- d. The Shift Supervisor or Emergency Director shall give pemission to re-open containment isclation valves.
- e. Installed radiation monitors or portable monitors shall be available to dei.act any release that may result from opening the valve.
j 4
4 4
9 i
1 l
J 13.0 FOR USE IN UV-~ l Oh LY .
.FOR USE IITUNITTUWLU.
1202-6B Revision 3 06/30/80 TABLE #1 HPI Actuation A (left) Actuation B (Right)
Status Status Status Light Description Light Status Light De.r:ription Light Color Color A. BWST SUP MU-V-14A Blue A. BWST SUP MU-V14B Blue B. R.B. SUMP DH-V6A Blue B. R.B. SUMP DH-V6B Blue C. DH to MU DH-V7A Blue C. DH to MU DH-V7B Blue D. Diesel A Running Blue D. Diesel B. Running Blue E. Aux. Bldg. MCC A Bkr Blue E. Aux. Bldg. MCC B Skr Blue F. Scrn. House MCC A Bkr Blue F. Scrn. House MCC B Bkr Blue G. LPI Line DH-V4A Blue G. LPI Line DH-V4B Blue H. BWST SUP DH-V5A Blue H. BWST SUP DH-V5B Blue I. R.B.C. Pump A. F.-P1A Blue I. R.B.C. Pump B RR-P1B Blue Amber
~~
J. RBC PP REC RR-V10A Amber -J. RBC PP REC RR-V10B--
K. RB Fan A AH-E1A Blue K. RB Fan B AH-ElB Blue L. RB Fan Mtr NS-V52A Blue L. RB Fan Mtr NS-V52B Blue M. RB Fan Mtr NS-V53A Blue M. RB Fan Mtr NS-V53B Blue N. DR Pump A DR-P1A B1 e N. DR Pump B DR-P1B Blue
- 0. DR Pump Dis DR-V1A Blue 0. DR Pump Dis DR-V1B Blue P. NR Pump A/B NR-P1A/B Blue P. NR Pump B/C NR-P1B/C Blue Q. NR Pump Dis NR-V1A/B Blue Q. NR Pump Dis NR-V1B/C Blue R. D.C. Pump A DC-P1A Blue R. D.C. Pump B DC-P1B Blue S. NS Pump A/B/ NS-P1A/B Blue S. NS Pump B/C NS-P1B/C Blue T. NS Deicing NR-V4A Blue T. NS Deicing NR-V4B Blue U. Aux. BL Fan AH-E-15A Blue U. Aux. BL Fan AH-E-15B Blue V. SCR HS Fan AH-E-27A Blue V. SCR HS Fan AH-E-27B Blue W. H-P Recire MU-V36 Blue W. H-P Recire MU-V37 Blue 14.0 FOR USE IN UNIT I ONLY
.. N WJGA5 UUVM
. , 1202-6B Revision 3 06/30/80 TABLE #1 HPI (Cont'd)
Actuation A or B (Center)
Status -
Status Ligh Description Light Color -
A. Non ESS Dump NS-V32 Blue B. RB Fan C AH-E1C Blue C. RB Fan Mtr NS-Y52C Blue D. RB Fan Mtr NS-V53C Blue i
-- ~ -
l 1
1 l
l l
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l l 15.0 ,
FOR USE IN UhlT I ON LY l 1 . . ,
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- 96/30/80 --
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==. .
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. ACCEPTABLE PRdkSURh - TEMPERATURE ~~-- - f , _ _- :- -_-_ _ _ -- ' -
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Each Div = 2.5 ,F
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FU M U d t- IIN U W1 i i viu i '~
1202-6B Revision 3 06/30/80 ATTACHMENT 1 RESTART OF REACTOR COOLANT PUMPS FOLLOWING AUTOMATIC ESAS ACTUATION 1.0 pREREOUISITES NOTE: The normal limits for minimum RC pressure for RC pump operation are superseded by these instructions. RC pump shaft vibration may also exceed normal limits. If the vibration is off-scale (>40 mils) the pumps should only be run if incore Tc temperatures are greater than Curve 2 Figure 3.1 of Attachment 3.
1.1 Restart of the Reactor Coolant pumps is recomnended for the ,
following conditions:
- a. Feedwater flow is available (steam generators at approximately 95".) and RCS repressurizes to greater than 1600 psig.
Refer to Section 2.1 of Attachment 1.
- b. Feedwater flow is available and RCS pressure exceeds secondary system pressure by 600 psig, refer to Section 2.2 of Attachment 1.
- c. Secondary system pressure is less than 100 psig and RCS l
[
pressure is greater than '250 psig, refer to Section 2.3 I i
of Attachment 1.
- d. Inadequate core cooling, refer to Attachment 3.
- e. Non-LOCA - Overcooling Event, refer to Attachment 2.
1.2 Verify the following RC pump services.
- a. Seal injection total flow >32 gpm. l
- b. Seal cooling: Intermediate Closed Cooling pump on, IC-V2, 3, and 4 open.
.FORQ5t IKUWFFFUIC 1202-6B Revision 3 )
06/30/80 1
- c. RC Pump Motor Cooling, NS-V4, 15, and 35 open. ;
1.3 Start RC Pump Motor Backstop 011 Pump.
1.4 Start RC Pump Motor Oil Lift Pump.
2.0 RESTART-OF RC PUMP 2.1 RCS Repressurized to >1600 osig.
- I
- a. Verify that emergency feedwater is available and OTSG 1evels are approximately 95% in at least one steam generator.
- b. Start one RC pump on the operable steam generator.
- c. Continue cooldown at approximately 1000 F/ hour but within Technical Specification limits.
2.2 RCS Pressure Exceed Secondary System Pressure by 600 osig.
- a. Verify that emergency feedwater is available and OTSG levels are approximately.95% on at least one steam generator.
I
- b. Bump one RC pump on the steam generator with feedwater for 10 seconds. ,
NOTE: Bumping of RC pump on steam generator with low steam generator level would not help condense the trapped RC vapors.
- c. Allow RCS pressure to stabilize while continuing to cool down. l I
- d. If after 15 minutes, RCS pressure exceeds secondary system 1
pressure by 600 psig bump a different pump on steam generator .
with feedwater for 10 seconds. Bump alternate RC pumps so that no pump is bumped more than once in an hour.
- e. After five (5) bumps allow the RC pump to continue in operation.
2.3 Secondary System Is less Than 100 osig and RCS Pressure Is Greater
'Than 250 osig.
19.0 FO R USE I \1 UN IT I O N LY
- 9
' . FT.)ffU5t IWU WFI I UIN L I l l
1202-6B Revision 3 06/30/80
- a. Verify that emergency feedwater is available and steam generator ,
is available and steam generator level is approximately 95% on ,
at least one steam generator,
- b. RC pressure has been stabilized for greater than one hour. )
- c. Bump ar. RC pump on steam generator with feedwater for approximately I1 1
10 seconds. i
- d. Wait 30 minutes and start an alternate RC pump, if RC flow is
'i indicated and established, maintain OTSG cooling by adjusting !
turbine by pass valves for the respective OTSG.
j l
1 I
20.0 l FO R USE IN Uh r I ON LY
FOR USE IN U NI I FONI_ T 1202-6B .
Revision 3 06/30/80
^
ATTACHMENT 2 NON-LOCA OVERC00 LING TRANSIENT WITH FEEDWATER AVAILABLE ,
- 1. If a steam line break is indicated, verify the following valves closed for the affected steam generator.
A Steam Generator B Steam Generator i FW-V17A W-V17B FW-V16A FW-V16B FW-V92A FW-V92B FW-V5A FW-V5B EF-V30A EF-V30B MS-V3D MS-V3A MS-V3d MS-V3B MS-V3F MS-V3C Refer to AP 1203-24 Steam Supply System Rupture.
- 2. If an overfeeding event is indicated close the following i valves for the affected steam generator.
4 A Steam Generator B Staam Generator FW-V16A FW-V16B FW-V17A FW-V17B FW-V5A FW-V5B FW-V92A FW-V92B EF-V30A EF-V30B If unable to isolate main feedwater flow trip the main feedwater pumps, i
- 3. InTaediately restart one RC pump in each loop if the RCS is ;
50 F subcooled. j I
21.0 FOR USE IN Uhl- l ON LY
' FOR USE IN UNITTUNI r "'
- - - 1202-6B Revision 3 05/30/80
- 4. Control steam pressure via turbine bypass or atmospheric dumo ,
valves to stabilize or control plant heatup. ,
NOTE: ' Considerable HPI may have been added to the RCS.
Therefore, to prevent RCS from going solid, it may be necessary to reduce secondary steam pressure.
I If the RCS becomes solid recover from solid opera-tion per OP 1103-5 enclosure III.
- 5. As long as the' RCS is maintained 50 F subcooled bypass ES and throttle HPI/MU and letdown flow to maintain pressurizer level at %100 inches.
- 6. Using turbine bypass valves and feedwater system, control steam generators as needed to limit plant heatup until RC pressure control can be re-established with the pressurizer.
NOTE: Cold RCS water MAY have been added to the pressurizer;
' therefore, a period of time may elapse before normal RC pressure control can be established with the pressurizer heaters.
- 7. Once pressure control is re-established, use normal heatup/cooldown procedure to establish desired plant conditions.
22.0 FO R USE IN UNIT I ON LY 5
.1-U H U d t- IN U lN I I I V I N L. I
. 1202-6B R: vision 3 06/30/80 ATTACHMENT '3 ,
ACTIONS FOR INADEOUATE CORE COOLING NOTE: Where instructions require opening the PORY, it must be accom-plished by inserting key into the PORY manual control key ,
switch and keeping the spring loaded switch in open position for the duration PORY needs to be open.
1.0 IMMEDIATE STEPS FOR INADE00 ATE CORE COOLING.
NbTE: If RC pumps are mnning, do not trip pumps. This supersedes instructions in Section B Manual Action. ,
1.1 Yerify HPI/LPI systems are functioning properly with maximum flow. Start third HPI pump and open MU-V217, if possible, to increase injection flow.
1.2 Verify steam generator level is being controlled at 95% on operate range.
1.3 Depressurize operative steam generator (s) to establish a 1000 F/ hour decrease in secondary saturation temperature.
1.4 Ensure core flood tank isolation valves CF-V1A and B are open.
1.5 If reactor coolant system pressure increases to 2300 psig open pressurizer PORV to reduce reactor coolant system pressure.
Reclose PORY when RCS falls to 100 psig above the secondary pressure. Repeat if necessary. If PORV is not operable, pressurizer safety valves will relieve pressure.
1.6 When the indicated incore themocouple temperatures or hot leg RTD temperatures are superheated for the existing RCS pressure, operator action shall be based on conditions determined from Figure 3-1, by a sample of the highest incore thermocouple temperature readings to detemine the, core exit themocouple temperature.
2 FOR USE IN tSNIT I Oh LY .
.FOR USE IN UNITTONLT 1202-6B l
- - Revision 3 1 06/30/80 )
l NOTE: More than one themocouple temperature reading should be used (for example use the average of 5).
1.7 When the incore themocouple temperature has been determined, proceed per the following guidelines.
INCORE THERMOCOUPLE TEMPERATURE SECTION 6.B.3 Follow-Up Action Incore Tc 1 Saturation Incore Tc > Saturation but less than Curve 1 Repeat Section 1 Curve 1 1 incore Tc < Curve 2 Figure 3-1 2 Incore Tc > Curve 2 Figure 3-1 3
NOTE: The incore thermocouple temperature readings shall be continuously monitored until the indicated i,ncore themocouple temperatures return to saturation temperature for the existing RCS pressure.
2.0 ACTIONS FOR CURVE 1 1 INCORE Tc < CURVE 2 FIGURE 3-1.
2.1 If RC pumps are not operating, start one pump per loop (if possible). This instruction supersedes previous instructions to trip RC pumps.
NOTE: Do not bypass nomal interlocks. Operate RC pumps until pump shaft vibrations are up to 30' mils.
2.2 Depressurize operative steam generator (s) as rapidly as possible to 400 psig or as far as necessary to achieve a 100 F decrease in secondary saturation temperature.
c 2.3 Open the PORV, as necessary, to maintain RCS pressure within 50 psi of steam generator secondary side pressure.
24.0 FOR USE IN UN'T I ONLY
M M U Db lW UINII I VINw I '
1202-6B , .
f R: vision 3 06/30/80 l l
l NOTE _: If steam generator d'epressurization was not possible, j
~
open PORY and leave open.
0 2.4 Immediately continue plant cooldown by maintaining 100 F/ hour.
Decrease secondary saturation temperature to achieve 150 psig g RCS pressure.
CAUTION: If emergency feed pump EF-P1 is supplied by main steam, do not decrease steam pressure below 150 psig, necessary for EF-P1 operation.
2.5 When RCS pressure reaches 150 psig, proceed per section 6.B.4
'Long Tem Cooling'.
2.6 If the average incore themoccuple temperature increases to Curve 2 Figure 3-1 proceed imediately to Section 3.0.
3.0 ACTIONS FOR INCORE Tc > CURVE 2 FIGURE 3-1.
3.1 If possible, start all RC pumps.
I l NOTE 1:
Starting interlocks may be defeated if necessary.
NOTE 2: High levels of shaft vibration should be expected.
The RC pumps should not be tripped if vibration limits are exceeded. _ __
3.2 Depressurize the operative steam generator (s) as quickly as possible to atmospheric pressure. (150 psig if EF-P1 is in operation using main steam).
CAUTION: If emergency feed pump is supplied by main steam, do not decrease steam pressure below 150 psig. a b
1 3.3 Open the pressurizer PORV and leave open. A e NOTE: The RCS will depressurize and the LPI system should restore core cooling. _
25.0 FOR USE IN UNIT l ONLY
. t-U M U d t liv uTvri cun
~
.1202-6B i Revision 3 06/30/80
] /
3.4 When incere themoccuple temperatures return to the saturation temperature for the existing RCS pressure - and - the LPI ,
system is delivering flow, proceed as follows: ,
3.4.1 Close the pressurizer PORY; reopen if RCS pressure increases above 150 psig.
3.4.2 Decrease to two (2) RC pump operation (one per loop).
3.4.3 Isolate the core flood tanks.
3.4.4 Maintain steam' generator pressure at atmospheric pressure (150 psig if EF-Pl is in operation using main steam).
3.4.5 HPI flow may be throttled if criteria c'f 6.B.3.12 is met.
3.4.6 Monitor BWST level as Lo-Lo level limits are approached, align LPI system for suction frcn RB sump per Section 6.8.4 Long Tem. -
NOTE: If HPI is required per 6.B.3.12 align LPI and HPI in piggyback mode. Close HPI suction valves to BWST.
f
{
.h 26.0 FOR USE IN UNIT I ONLY -
,, N DTQPQ C WW LVJTNIhTm.UUVw I 5MwMUMt.- rC umuwur er--- - ------
06/30/30
- .. INADE00 ATE CORE COOLING
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F.U n UUC IIN UTNN
- ' 1202-6B 1 Revision 3 1 06/30/80 ATTACHMENT 4 -
ACTIONS FOR FAILURE OF THE EMERGENCY FEEDWATER SYSTEM A. Failure of EF-P1 to Start.
- 1. Verify that MS-V2A and MS-V2B are open.
- 2. Verify that CO-V10A, CO-V10B, EF-V1A and ZF-V1B are open.
- 3. Press the OPEN push button for MS-V13A and MS-V138.
4 If EF-P1 fails to start press the CLOSE push button for MS-V13A and MS-V13B.
- 5. Have an Aux. Operator check locally that EF-P1 overspeed trip
- is reset and that the Manual Operator for MS-V6 is in the open position.
B. Failure of EF-P2A or EF-p2B To Start. .
- 1. Verify that there is voltage available at the associated bus.
NOTE: With an ES signal present there will be a 20 second delay in pump start.
- 2. Verify that control power is available as indicated by the green indicator light at the control switch.
- 3. Manually start the pump using the control switch.
NOTE: If EF-P1 and either motor driven pump have started any. further investigation may be perfomed when plant conditions become stable.
- 4. Check locally for targets on relays located at the switchgear.
- 5. Use the 69 bypass key and attempt to start the pump locally.
C. Failure of EF-V30A or EF-V30B to open.
- 1. Check steam generator level to detemine whether an open signal is required. The valves should be open if steam generator l 1evel is < 30 inches on the startup range (RC Pumps on) or <
50t on the operating range (RC Pumps off).
FOR USE IN hN T I ONLY -
l
- 1202-6B .
Revision 3 06/30/80 .
- 2. Verify that EF-V30A and EF-V30B Hand-Auto Stations are in AUTO.
- 3. If the valves have failed to open, place the Hand-Auto Station in RAND and attempt to open the valve.
- 4. If the valves have not opened, attempt to open using the backup Manual Control Station located in the control room.
4
- 5. If the valves are still not open, have the aux. operator establish communication with control room operator and take local handwheel control of the valves and open them as directed by the control room operator.
D. No Indicated Flow (Low Flow) With Steam Generator Level Below~
Setooint.
- 1. Verify that EFP discharge pressure is higher than steam generator pressure. If not Turbine Header pressure setpoint may be reduced to get additional flow.
- 2. Verify that the following valves are open:
EF-V2A EF-V30A EF-V2B AND .
CO-V10A EF-V30B -
CO-V10B
- 3. Have aux, operator check locally for pump malfunction and correct lineup of manual valves.
W 29.0 FOR USE IN U NIT I ON LY
rurruvi rr v rivr -
1202-6B
, RCS Leak /Depressurization
' ~
3 EP 1202-6A, 6B and 6C $N$)so ATTACHMENT 5 NO is RCS Press
<1600 psi e i YES ,
Verify Reactor tnp Verify HPl .
Trip RC Pumos YES 1202 6A is LPI flow initiated (<200 ps0 l NO Increase RCS MU, Shutdown Reactor 1202 6C 12026B Verify main or emerg .
Verify RB isol FW flow & level '
Verify EFW y,h k'*3p Did RCS Press YES , Verify LPI s 1600 psi is event Non LOCA YES Overcoormg fj0 1202 6B l NO
- Attachment 2 ..
YES OTSG Tube Rupt
_ YES Did RCS 1202 5B NO ' < Repressurize Verify Cent isol NO NO is margin from ,
saturation > E0 *F l Are symptoms YES
->- of inadequate cara initiate full (
YES cooling present HPI and verify 1202 6B OTSG cooling i NO Attachment 3 NO is saturation is N at C re present ,
reached with YES Verity. OTSS temp increase YES Cooling. continue l NO Cooldown at fuH HM Continue HPI ~100 *Fihour BWSTLola Until 50 ' margin ' YES is 50'F margm. Level Alarm regained gp3
, from sat. available revent solid cend YES l NO Cooldown per :
NO !s LPl>1000 gpm OP 110211 18.RC, Pp restart '
Bump (Start) entenamet Attach 1 for>20 min RC Pumps Estab HPllLPI from RB Sump YES BWST La Lo Level Estsbkh LPI Alarm Flow fromRB Sump NO is LPI >1000 gpm: YES for 20 min !
Establish HPilLPI Estabbh LPI Flow frcm RB Sump Flow from RB Sump FO R USE IN lJM- 1 ON _Y
. ... .. 9 e " **
j i
UNITED STATES OF AMERICA
- NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
)
In the Matter of ) '
)
METROPOLITAN EDISON ) Docket No. 50-289 COMPANY, et al.,
-) .
)
(Three Mile Island )
Nuclear Station, Unit )
No. 1) )
)
CERTIFICATE OF SERVICE I hereby certify that copies of the " Direct Testimony of Robert D. Pollard o'n Behalf of the Union of Concerned Scientists Regarding UCS Contention No. 5" have been mailed postage pre-paid this 10th day of October,1980 to the following parties:
Secretary of the Commission (3) Mr. Steven C. Sholly U.S. Nuclear Regulatory Commission 304 South Market Street Mechanicsburg, PA 17055 Washington, D.C. 20555 Attn: Chief, Docketing & Service Section
- James A. Tourtellotte, Esq. Jordan D. Cunningham, Esq.
- Of fice of the Exec. Legal Director Fox, Farr & Cunningham U.S. Nuclear Regulatory Commission 2320 North Second Street-Washington, D.C. 20555 Harrisburg, PA 17110 Karin'W. Carter, Esquire Frieda Berryhill
. Assistant Attorney General Coalition for Nuclear Power 505 Exedutive House Postponement .
P.O. Box 2357 .~ 2610 Grendon Driv,g' Harrisburg, PA 17120 Wilmington, Delaware 19.803 Daniel M. Pell Walter W. Cohen, Consumer Adv.-
,32 South Beaver Street Department of Justice -
York, Pennsylvania 17401 Strawberry Square, 14th Floor Harrisburg, PA 17127 i -
- - t s ,
Cert. of Service Docket No. 50-289 Robert L. Knupp, Esquire Chauncey Kepford Assistant Solicitor' Judith H. Johnsrud County of Dauphin Environmental Coalition on P.O. Bo.; P Nuclear Power i 407 North Front Street 433 Orlando Avenue l
Harrisburg, PA 17108 State College, PA 16801 .
John A. Levin, Esquire Robert Q. Pollard Assistant Counsel Chesapeake Energy Alliance Pennsylvania Public Utility 609 Montpelier Street Commission Baltimore, Maryland 21218 l Harrisburg, PA 17120 l
Theodore Adler Marvin I. Lewis Widoff, Reager, Selkowit: 6504 Bradford Terrace l
& Adler Philadelphia, PA 19149 l 3552 Old Gettysburg Road l Camp Hill,'PA 17011
~ ~ ~Ms. Marjorie Aamodt --Ivan W. Smith, Chairman RD #5 Atomic Safety & Licensing Board Coatesville, PA 19320 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dr. Walter H. Jordan Dr, Linda W. Little 881 W. Outer Drive 5000 Hermitage' Drive Oak Ridge, Tennessee 37830 Raleigh, North Carolina 27612
- George F. Trowbridge, Esquire Ms. Jane Lee l R.D. #3, Box 3521
! Shaw, Pittman, Potts &
Trowbridge Etters, Pennsylvania 17319 1800 M Street, N.W. .
20036 l
Washington, D.C.
j j ~
w- -
- Hand delivered .
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