ML19351F125

Testimony Re Ucs Contention 7.Absence of Direct Indication of Core Water Level Presents Undue Risk to Public Health & Safety.Prof Qualifications & Certificate of Svc Encl

ML19351F125
Person / Time
Site:	Three Mile Island
Issue date:	12/04/1980
From:	Pollard R UNION OF CONCERNED SCIENTISTS
To:
References
NUDOCS 8012290547
Download: ML19351F125 (22)

Text

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a UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEF0FE THE ATOMIC SAFETY AND LICENSING BOARD

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In the Matter of )

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METROPOLITAN EDISON ) Docket No. 50-289 CCMPANY, et al., )

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(Three Mile Island ) <

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'sg '.i 4 DIRECT TESTIMONY OF -M ROBERT D. POLLARD ON BEHALF OF THE UNION OF CONCERNED SCIENTISTS REGARDING UCS CONTENTION NO. 7 e

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! ROBERT D. POLLARD i OUALIFICATIONS I

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 professionals supported by over 80,000 public sponsors.

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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 training, 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.

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After his honorable discharge in 1965, Mr. Pollard i attended Syracuse University, where he received the degree of Bachelor of Science maana cum laude in Electrical Engi-i 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 1,

at the Graduate School of the University of New Mexico in Albuquerque. He subsequently advanced to the cositions of Reactor Engineer (Ins trumen t a tion ) and Project Manager with

, AEC/NRC.

As a Reactor Eng i nee r , 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 i 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, Uni t 3, Comanche Peak, Units 1 and 2, and Catawba, Units-1 and 2. While with NRC, Mr. Pollard also served on the standards group, participating in developing standards and safety guides, and as a member of IEEE Committees.

OUTLINE - DIRECT TESTIMONY ON UCS CONTENTION NO. 7 The testimony provides examples showing that the TMI-2 accident demonstrated the hazards that can result when plant instrumentation does not, in an unambiguous way, provide the information needed to determine the course of action necessary to protect public health and safety.

IEEE Std, 279, incorporated into 10 CFR 50.55a(h) contains requirements calling for direct measurement of desired variables and avoidance of anomalous indications. In the case of inadequate core cooling, the desired variable is water level in the reactor vessel. Rather than prcvide

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instrumentation to directly measure water level, Met Ed proposes to rely on indirect indications. The testimony-discusses the defici ncies in the current and proposed TMI-l instrumentation and concludes that the absence of a direct indication of core water level presents undue risk to public health and safety.

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UCS CONTENTION NO. 7 NRC regulations require instrumentation to monitor variables as appropriate to ensure adequate safety (GDC 13) and that the instrumentation shall directly measure the desired variable. IEEE 279, S4.8, as incorporated in 10 CFR 5 0. 55a (h) , states that:

To the extent feasible and practical protection system inputs shall be derived from cignals which are direct measures of the desired variables.

TMI-l has no capability to directly measure the water level in the fuel assemblies. The absence of such instrumentation delayed recognition of a low water level condition in the-reactor for a long period of time. Nothing proposed by the Staff would require a direct measure of water level or provide an equivalent level of protection. The absence of such instrumentation poses a threat to public health and safety.

The TMI-2 accident demonstrated the wide range of serious safety hazards that can result when the plant instru-mentation does not provide the information needed to determine the course of action necessary to protect public health and safety. The lack of instrumentation to measure auxiliary

7-2 feedwater flow contributed to the delay in recognizing that there was no auxiliary feedwater flow. The inability to distinguish the personnel exposure to radioiodine from that of noble gases caused the operators to believe that respiratory protection was needed, thereby unnecessarily hampering their performance and communications. The absence of instrumentation to determine whether the PORV was open or shut prevented early recognition that a loss of coolant accident was in progress.

The lack of an instrument to measure reactor water level contributed to the fallare to recognize that the fuel was uncovered.

The Commission's regulations contain requirements that, if enforced, provide a high degree of assurance that such problems will not arise. 10 CFR 50.55a (h) requires that protection systems meet the requirements of IEEE Std 279,

" Criteria for Nuclear Power Plant Protection Systems."

Section 4.8, " Derivation of Systems Inputs," requires that

"[t}o the extent feasible and practical, protection system inputs shall be derived from signals which are direct measures of the desired variables." Section 4.20, "Information Read-Out,"

requires that "[t]he pro ~tection system shall be designed to provide the operator with accurate, complete, and timely infor-mation pertinent to its own status and to plant safety. The

7-3 design shall minimize the development of conditions which would cause meters, annunciators, recorders, alarms, etc.,

to give anomalous indications confusing to the operators."

In addition to applying to protective actions that are per-formed automatically, the requirements of IEEE Std 279 must also be applied to the equipment used by the operator. The appendices to Sections 7.2 and 7.3 of the Standard Review Plan state that the requirements of IEEE Std 279 apply to all equipment used by the operator to detect the need for protective action, to acccmplish the protective action, and to confirm completion of the protective action.

The requirements of IEEE Std 279 calling for direct measurement of the desired variables and for avoiding the development of conditions that would give anomalous indications confusing to the operator are reflected in some, but not all, of the Positions in NUREG-0578. For example, " Safety grade indication of auxiliary feedwater flow to each steam generator shall be provided in the control room." (Page A-32); "Each licensee shall provide equipment...for accurately determining the airborne iodine concentration throughout the plant under accident conditions." (Page A-41); and " Reactor system relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve position detection device or a reliable indication of flow in the discharge pipe." (Page A-10). In the case of the

7-4 last example, NUREG-0578 cites the requirement of IEEE-279 for direct measures of the desired variable as the basis for requiring "a more positive indication of valve position."

(Page A-9). This is intended to correct the pre-accident situation when only " indirect indication of safety [and]

relief valve position" was available - a situation that was

" misleading" and " ambiguous." (See NUREG-0578, pages A-9 and A-10).

In ccntrast, the Position of NUREG-0578 concerning instrumentation for detection of inadequate core cooling does not require conformance with the Commission's requirement for direct measures of the desired variable. The TMI-2 accident involved low water level in the reactor pressure vessel. The operator did not recognize this condition (and no safety grade systems were provided to detect and correct this condition). In my opinion, the principal factor that prevented the operator from recognizing that there was a low water level in the reactor was the absence of instrumentation to directly measure that water level. However, the position in NUREG-0578, in apparent disregard of the Commission's regulations embodied in:IEEE-279 and GDC-13, is based on the following assertion: "With the hindsight of TMI-2, it appears that the as-designed and field-modified

• instrumentation

• This phrase apparently means modified during the accident.

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at Three Mile Island Unit 2 provided sufficient information to indicate reduced reactor vessel coolant level, core I voiding, and deteriorated core therral conditions." (NUREG-0578, page A-ll). In other words, despite the absence of a direct measure of reactor water level and the fact that the i

operator failed to recognize the low water level condition, the operator could have recognized that condition by deduc-tions based on the indirect information available. The fact is that the reactor operators and their supervisors, acting

! on the basis of the instrumentation available to them, failed to discern or to believe that the core was uncovered for a long-time.  ;

Nevertheless, NUREG-0578 concludes that the problem j can be addressed in two stages. "The first stage is based on the detection of reduced coolant level or the existence i

of core voiding with the existing plant instrun.entation."

(NUREG-0578, page A-11. ) The second stage is to study and develop system modifications to provide a more' direct

. indication of low reactor water level or inadequate core cooling than that available with present indication. The i

changes to be studied include reactor vessel water level detectors for pressurized water reactors. However, the study f is to be restricted to modifications "that would not require 4

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7-6 major structural changes and that could be implemented in a t

relatively rapid manner." (NUREG-5078, page A-11). The i Staff does not define " major" and "relatively rapid" nor does it disclose any basis for these restrictions on the study.

However, in my opinion, the provision of reactor vessel water level detectors does not necessarily involve " major structural modifications." Furthermore, the Commission's decisions are supposed to be based on safety considerations i

rather than on whether the necessary safety provisions are expensive or time-consuming.

I conclude that the Position in NUREG-0578 concerning instrumentation for detection of inadequate core cooling will not achieve the stated goal of providing " substantial, addi-tional protection required for the public health and safety."

(NUREG-0578, page 3). One of the principal lessons to be i learned from the TMI-2 accident is that the health and safety of the public is subject to undue risk when instrumentation to directly measure those plant variables needed to assure safety is not provided. Indeed, the Staff has acknowledged that direct measurement of reactor vessel water level would provide important information during a small break loss-of-l coolant accident. "A direct measurement of reactor coolant

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l j level would be of assistance to the reactor operator: in

7-7 diagnosing a small break LOCA; in determ!.ning that the I course of events was as expected for a small break LOCA; and in determining the effectiveness of autcmatic and manual actions." (Response to UCS Interrogatory 68) . As i

i late as March 31, 1980, the Staff position was that "[w]ater i level indication is considered...to provide a more direct i

measuremer.t of reactor vessel level and the licensee has been required to consider this indication in his evaluation of additional ingtrumentation." (Response to UCS Interro-gatory 73). To argue that instrumentation which does not directly measure reactor water level and which proved inadequate during the TMI-2 accident, can be relied on to detect low.

level in the future contradicts the lesson to be learned. .

Substantial, additional protection for the public can be proviced by instrumentation that directly measures reactor vessel water level. I agree with the Staff's statement that "[n]one of the instrumentation proposed for TMI-l for i

other purposes would provide a reliable indication of reactor coolant level under all conditions." (Response to UCS Interrogatory 74). Therefore, I conclude that TMI-l should not be permitted to restart. I also conclude that the study Met Ed is required to conduct to develop system modification to provide more direct indication does not now provide any .

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7-8 protection to the public. Therefore, even " reasonable progress" on that study cannot form the basis for concluding that the Commission's regulations are met or that restart will not pose undue risk to public health and safety.

I will now address the other deficiencies in the Staff's and Met Ed's positions which are not explicitly addressed above.

Met Ed acknowledges that reactor level information may be of assistance to the operator (Response to UCS Interrogatory 66), but notes that reactor water level instrumentation is not required by the short term and/or long term measures in NUREG-0578 or NUREG-0585. (Response to UCS Interrogatory 68).

As to conformance with Section 4.8 of IEEE-279, Met Ed argues that since the safety analyses in the TMI-l Final Safety Analysis Report do not require a reactor water level signal as an input to the plant protection systems, IEEE-279 does not apply. (Response to UCS Interrogatory 69.) In my view, this is tantamount to arguing that since the plant was designed to rely on an indirect measurement of the desired variable, the Commission's regulation requiring a direct measurement is not applicable. Clearly the conclusion that should be drawn, especially in view of the TMI-2 accident, is that the Commission's regulation is not met. The Staff advances a different

7-9 justification for concluding that "the inclusion of indirect measurement level indication does not, in the Staff's view, constitute a violation of Section 4.8 of IEEE 279-1968."

(Response to UCS Interrogatory 67, emphasis added) . The Staff's principal justification is that the direct measurement of reactor vessel water level "was not conclusively known to be feasible or practical." (Ibid). The Staff can hardly rely on technological ignorance when direct measurement instru-mentation is used to indicate water level in pressurizers, steam generators and boiling water reactor pressure vessels.

As an additional excuse for not requiring direct measurement level instrumentation, the Staff made the statement that Met Ed and B&W, not the NRC, should determine what instrumen-tation is appropriate: "The applicant and reactor vendor are in a better position to assess the instrumentation bGst suited to determine water level within the vessel for their

plant." (Response to UCS Interrogatory 67) . I trust this Board will not be similarly inclined to leave it to Met Ed to l

determine whether TMI-l meets the Commission's regulations.

l Met Ed argues that the addition of instrumentation to l indicate either the pressure or temperature margin to saturation in the reactor coolant loops, the expanded range of the hot leg loop temperature instruments, and the incore thermocouples "will provide more direct measurement of reactor cooling level l

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and inadequate core cooling." (Response to UCS Interrogatory 72). The Staff lists the same instrumentation as Met Ed and adds the source range neutron detectors, reactor coolant pump motor current and reactor coolant flow as additional possible indications of inadequate core cooling. (Response to UCS Interrogatory 69). My conclusion is that none of the instruments identified by Met Ed and the Staff provide a direct indication of reactor coolant level or inadequate core cooling. In fact, those instruments provide essentially the same information as that available prior to the TMI-2 accident. I note the Staff's agreement with this conclusion:

"However, we note that the instrumentation at the [TMI-1],

facility remains fundamentally unchanged and is still similar to that at TMI-2 at the time of the accident." (TMI-l Restart Evaluation, page C8-16).

The current instrumentation is inadequate and/or potentially

. misleading for the following reasons:

1. By the time the source range neutron detectors indicate increased neutron leakage, there must already be a condition of low vessel water level

! and inadequate core cooling. Therefore, these instruments do not allow the operator any advance warning of potentially dangerous conditions. Further-more, increased neutron level is not an unambiguous

7-11 indication of inadequate core cooling. It could instead indicate that the reactor was critical, as the TMI-2 operators apparently believed.

2. Reactor coolant pump flow and motor current in-dications cannot be relied upon because:

a) the reactor coolant pumps will not be operating if offsite power is lost, and bl new post-TMI emergency procedures call for the reactor coolant pumps to be shut off if ECCS was initiated by low reactor coolant pressure.

3. The hot leg temperature indicators are not loca.ted in the reactor vessel and they measure the temperature of whatever material is present at theirlocation, be it water or gas. Thus, the accumulation of hydrogen or nitrogen (from the core flooding tanks) could not only stop natural circulation, but also give a false indication of the temperature in the i

vessel. Furthermore, the measurement of temperature l

in the hot legs is of no use during the " bleed and feed" mode of core cooling because it is not in the cooling water flow path.

4. The saturation meters have the same deficiencies as the hot leg temperature indicctors because they

7-12 utilize the same input signal. The saturation meters have two additional deficiencies. First, with only one meter per loop, the design does not meet the single failure criterion. If heat removal is being achieved through only one loop'and steam generator, as it was during part of the TMI-2 accident, there is no backup if the one saturation meter in that loop fails.

Second, the saturation meters are designed such that there is a significant potential for confusing the operator, particularly in the stressful environ-ment of an accident situation. Procedures call for I the operator to maintain a 50*F temperature margin to saturation. However, the meter can display either temperature or pressure margin. If the operator makes the simple mistake of thinking the temperature

.~.argin is displayed when in fact the pressure margin

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is displayed, he could well fail to maintain an adequate margin to boiling. This can occur because the relationship saturation temperature and saturation pressure is not linear. For example, at 212*F, a pressure approximately 22 psi above saturation pressure is sufficient to achieve a 50*F margin to I

7-13 boiling. At higher. temperatures, maintaining reactor pressure only 22 psi above saturation would result in significantly less temperature margin.

For example, at 510*F, a reactor pressure 22 psi above saturation would leave less than a 3*F margin to boiling. To maintain a 50*F margin at 510"F requires that reactor pressure be approximately 350 psi above saturation pr ssure.

Despite these deficiencies in the present TMI-l instru-mentation, the Staff finds the instrumentation acceptable.

The Staff apparently expects the operator to deduce reactor vessel water level from instrumentation that is essential.ly unchanged as a result of the TMI-2 accident and does not directly measure reactor water level on the basis of only 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> of training the operators on the "new" procedures and instruments. (Response to UCS Interrogatory 72). It is my view that the deficiencies in the TMI-l instrumentation are so serious that they preclude a finding that.the Commission's regulations are met or that the post-TMI-2 modifications l

provide the substantial additional protection which is required l

for the public health and safety.

I believe it is important to express at this point l

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in my testimony, my conclusion that the design deficiencies l

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7-14 at TMI-1 extend beyond the lack of a direct measurement of reactor vessel water level. The continued reliance on operator action to assure adequate core cooling is, in my opinion, unjustified and unsafe. The TMI-2 accident graphically

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demonstrated that, under accident conditions, human beings are prone to forget their training and/or misapply procedures.

Furthermore, due to enormous numbers of possible combinations and sequences of events during an accident, it is impossible to foresee all events that could distract and confuse the operator. Therefore, I conclude that instrumentation provided .

to directly measure reactor water level should be used to automatically initiate any necessary protective actions., The fundamental, underlying reason for this is the same as I have discussed elsewhere in my testimony for other protective actions and for preventing improper operator interference in the automatic initiation and completion of other protective actions. Reliance on operators to perform safety functions offers little if any benefit and presents significant risks.

Another deficiency in the TMI-l instrumentation design is the frilure to comply with several of the " Clarification" l

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l For example, see my testimony on UCS Contentions Nos. 3 and 4 l concerning the inadequacies of relying upon operator action to disconnect loads from and connect the pressurizer heaters to the onsite power supply.

See my testimony on UCS Contention No. 10.

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7-15 items listed on page C8-17 of the TMI-l Restart Evaluation concerning the saturation meter. Items 3 and 4 require safety grade temperature and pressure inputs. However, neither of the inputs meets the requirements for safety grade instru-ments. The Staff attempts to dismiss this problem by: 1) address-ing those portions of safety grade requirements that are alleged-ly met and ignoring those that are violated; 2) noting Met Ed's plans to upgrade the instruments at some unspecified time in the future; and 3) pointing to the computer calculation of saturation margin and connection of the incore thermo-couples to the computer as backups to the saturation meter.

(TMI-l Restart Evaluation, page C9-18 and Response to UCS Interrogatory 67.) My evaluation of these statements is as follows: 1) safety grade means meeting all the requirements applicable to components important to safety. Commission policy does not permit arbitrary picking and choosing among those requirements; 2) plans to meet safety grade requirements in the future do not constitute a basis for finding adequate protection for the public now and, therefore, do not provide a basis for permitting restart; and 3) the computer backup does not meet safety grade requirements and does not meet Clarification item 6.

Item 6 requires that if the plant computer is to be used, its availability must be documented. However, the

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7-16 Staff argues that since the computer is a backup to the saturation meter, this requirement need not be met. (TMI-1 Restart Evaluation, page C8-18) . The Staff's reasoning comes full circle. The saturation meters need not have safety grade inputs because the computer is a backup and the computer need not be safety grade nor its availability documented because it is only a backup to the saturation meters. How this arrangement is supposed to provide the substantial, additional protection required for public health and safety is not explained by the Staff. I conclude that no such protection is provided.

Finally, Clarification Item 7 requires that, in the long term, the instrumentation providing indication of coolant saturation conditions must meet the requirements of Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident." The Staff states that since Revision 2 of this Guide is currently under development, Met Ed will not be required to satisfy Clarification Item 7 at this time.

(TMI-l Restart Evaluation, page C8-18.) Revision 2 of Regulatory Guide 1.97 is being developed as a result of the lessons learned from the TMI-2 accident. The revision will consider degraded j

core cooling conditions to a greater extent than the existing version which was issued in August 1977. The revised Guide i

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7-17 will describe an acceptable method of complying with the Commission's regulations applicable to instrumentation needed to ensure safety during and following an accident.

I conclude that until the evaluation of the TMI-2 accident is sufficiently complete to determine what requirements must be met to ensure public health and safety and until those requirements are met, TMI-1 should not be permitted to restart.

The fact that the Staff is unable to specify what those require-ments are precludes the finding the TMI-l can restart without undue risk to public health and safety.

In summary, the lack of a direct measure of water level in the core was a substantial t:ontributor to the severity of the TMI-2 accident. Despite chis fact, none of the modifications proposed for TMI-l are sufficient to provide an unambiguous indication of inadequate core cooling. It is my opinion that the Commission's regulations are properly interpreted to require a direct measurement of core water level and that the absence of such instrumentation poses an undue risk to public health and safety.

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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 on Behalf of the Union of Concerned Scientists Regarding UCS Contention No. 7" have been mailed postage pre-paid this 4th day of December, 1980 to the following parties:

Mr. Steven C. Sholly Secretary of the Commission (3)

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.

Office of the Exec. Legal Director Fox, Farr & Cunningham U.S. Nuclear Regulatory Commission 2320 North.Second Street I l 20555 Harrisburg, PA 17110 e l

Washington, D.C. '

Frieda Berryhill.

Karin W. Carter, Esquire

Assistant Attorney General .. Coalition for Nuclear Power .

Postponement

-505 Exec'utiv'e House *

' 2610 Grendon Drive P.O. Box 2357 Wilmington, Delaware 19808

.Harrisburg, PA 17120 ,,

l Walter W. Cohen, Consumer Adv.

( Daniel M. Pell Department of Justice 32 South Beaver Street Strawberry' Square, 14th' Floor

York, Pennsylvania 17401 Harrisburg, PA- 17127

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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. Box P Nuclear Power 407 North Front Street 433 Orlando Avenue 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 Harrisburg, PA 17120 Theodore Adler Marvin I. Lewis Widoff, Reager, Selkowit 6504 Bradford Terrace

& Adler Philadelphia, PA 19149 3552 Old Gettysburg Road 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 Shaw, Pittman, Potts & R.D. #3, Box 3521 Trowbridge Etters, Pennsylvania 17319 1800 M Street, N.W. -

Washington, D.C. 20036 .

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