ML19295D348

Testimony of J Wermiel,W Jensen,E Lantz & B Boger Before ASLB Re TMI-1 Restart Hearing Board Question 6,emergency Feedwater Reliability.Supporting Documentation Encl

ML19295D348
Person / Time
Site:	Crane
Issue date:	10/10/1980
From:	Boger B, Jensen W, Lantz E, Wermiel J Office of Nuclear Reactor Regulation
To:
Shared Package
ML19295D349	List: ML19295D348
References
FOIA-80-515, FOIA-80-555 NUDOCS 8011050003
Download: ML19295D348 (18)

Text

'

O UNIE D STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATCMIC SAFETY AND LICENSING 3 CARD In the Matter of

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METROPOLITAN EDISOff COMPANY,

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Occket No. 50-2S9 et. al.

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(Three Mile Island Nuclear

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Station, Unit 1)

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NRC STAFF TESTNONY OF J. WER.MIEL, W. JENSEN E. LANTZ, AND 3. 30GER REGARDING i.uERGENCY FEECWATER SYSTEM RELIABILIT(

(30ARD CUESTION 6 (Tr. 2394-96) Emercency Feecwater Reliability Question 6a. Is a loss of emergency feedwater following a main feedwater transient an accideat which must be protected against with safety-grade equipment? Would such an accident be caused or aggravated by a loss of non-nuclear instrumentation, such as occurred at Oconee?

(Witness Wermiel)*

Response: No. The loss of emergency feedwater folicwi tg a main feedwater transient is not an accident which must be protected against with safety-grade equipment.

With respect to the Oconee Incident, a complete loss of all feedwater will not result from a failure in the integrated control system /non-nuclear instrumenta-

.tion (ICS/NNI), since modifications will be made to the EFW system (as described

• Note: Name in parenthesis indicates preparer of response.

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. in the T71I-l Restart SER, NUREG-0680, pages CS-35, 36 and 37) that will completely eliminate any intertie between the ICS/NNI and EW systems. These modifica-tions will upgrade the EW system to a fully safety-grade system. Pric-to installation of the fully safety-grade system, an Cconee type event may result in a mcmentary loss of ER. Mcwever, this situation would be detected by the operator througn the EM flow indicators and steam generator level indication which are separate frem the NNI as described in the TMI-l Restart SER, NUREG-G620. The operator can then take the necessary manual action ~, the control room to open the EW control valves and restore feed flow.

(Wimess Jensen)

In the unlikely event tnat both main and emergency feedwater cannot be restored, the coerator at TMI-l is instructed to actuate High Pressure Infec-tion as described in Inadequate Core Ccoling Emergency Procedure EP1202-39.

This action will provide adequate cooling to the core by feed and bleed if two nign pressure injection pumps are avi.ilaole. The high pressure injection system cperates independent of NNI/ICS.

Question 6b.

In what respect 'is the emergency feecwater system vulnerable to non-safety-grade system failures and to operator errors?

(Witness Wemiel)

Response: The emergency feedwater system cur-ently meets all requirements of a safety-grade system and is, therefore, not vulnerable to ncn-safety-grade system failures with one exception. A single failure in the non-safety-grade integrated control system could result in a loss of flow by closure of the ficw centrol valves.

Implementatic s of the safety-grace modification to tne automatic initiation design, wnich Wil eliminate the single failure

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, possibility, will correct this deficiency. In addition, a question has been raised by the licensee in Licensee Event Report 30-012/0IT-0 dated July 11, 1980, concerning the adequacy of the environmental qualification for the emergency feedwater system ccmconents in the event of a main steam line break in the Intermediate Building. This area is being reviewed in connection with IE Bulletin 79-018.

With regard to the vulnerability of the EFW system to operato errors, the EFW system is vulnerable to operator errors as is any safety-grade system.

For example, the operator may leave valves closed or turn off pumps. However, improved emergency and operating procedures coupled with operator training in these procedures (as described in the TMI-l Restart SER, NUREG-C680) should limit the possibility of operator error.

Question 6c. What has been the experience in other power plants with failures of safety-grade emergency feedwater systems, if they have such systems in other power plants?

(Witness Lant:)

Response

We have reviewed the Licensee Event Reports for plants with safety-grade emergency feedwater (EFW) systems. The available data for plants that are in commercial operation and that have safety-grade emergency feed-water systems show that in the vast majority of the cases, the failures that occurred did not defeat the functional capability of the system.

Data

. on EFW system success on demand is not maintained. However, it should be noted that all plants perform routine periccic EFW system surveillance testing in accordance with specific plant Technical Soecifications. Except for the following reported cases of ccamon cause and operator-induced failure, wnich resulted in an overall system failure, the system was found to be operable when inf tf ated. These are cases w*ere sufficient emergency feedwater was not available, although energency feedwater was not required it the time to cool tne reactor.

Plant Date LER#

Descriotion Ginna 12/14/73 73-01T The suction was lost on two emergency feedwater pumps during a startup test because of the lack of air vents.

Turkey point 3 05/08/74 74-01T During a start test two emer-gency feedwater pumps failed to start due to tight packing. A third feedwater pump started but tripped because of foreign matter in tne governor.

Kewaunee 1 11/05/75 75-01T During startuo operations resin beads clogged the strainer to all emergency feedwater cumos.

Haddam Neck 07/05/76 76-03L When the plant was in :ne startup mode, both emergency feedwater pumps were vapor bound due to a leaky check valve.

It should be noted that prior to commercial operation at the Davis Besse plant, there was one event in which the safety grade emergency feedwater system failed on demand.

D' g Question 6d. What operator action is reauired to operate in a feed-and-bleed mode following a loss of emergency feedvater?

(Witness Boger)

Response: According to 5P 1202-26A, Loss of Steam Generator Feedwater to Soth OTSGs, if both main and emergency feedwater are lost, the ocerator is directed to initiate High Prassure Injection (HPI). HPI is actuated by depressing two pusnouttons on the main control board. The next operator action is to verify starting of the HPI pumps and opening of the HPI discharge valves. The operator is then instructed to verify that the PORY block valve is open and that the p0RV is cycling to maintain RCS pressure at approximately 2450 psig. At this point the operator has provided a maketp supply to the RCS (feed) and a relief path to remove core heat (bleed). If the 90RV or PCRV block valve fail to respond (open), the RCS. safety valves will relieve at 2500 psig to provide tne bleed path.

Suosequent operator actions include throttling the HPI discharge valves to maintain at least a 50*F margin of subccoling and attempting to restore a feecwater supply to the steam generators from the main or emergency feedwater system or the condensate system.

Question 6e.

If the emergency feedwater system were to fail, what assurance to we have that the system can be cooled by the feed-and-bleed mode? This is of particular concern if the PORVs and safety valves have not been tested uncer wo-ohase mixtures.

c

. (Witness Jensen)

Response: The answer to this question is contained in the NRC response to UCS Contention 1, questions 15,16, and 17.

Question 6f. Can the system be taken to cold shutdcwn with the feed-and-bleed cooling only? Are both high pressure injection (HP!) pumos required to dissipate the decay heat in the feed-and-bleed mode? The board would like an evaluation of the reliability of the feed-and-bleed system. Has tnere been any experience using that system?

(Witness Jensen)

Response: Analyses by Babcock _and Wficox indicate two high pressure injection pumos are required to adequately cool the core by feed-and-bleed for tne first three hours if decay heat is calculated using 1.2 times the AMS-5 decay heat model as required for LOCA analysis in Appendix K to 10 CFR 50. After tnree hours only one pump would be required. If a best estimate decay heat model is utili:ed (1.0 times the ANS-3 decay heat model) the analyses indicate that only one HPI pump would be required. See letter from J. Taylor, S&W, to R. "a ttson, NRC,."ay 12, 1979, wnich transmits Volume i Section 6.0 - Suoplements 1 and 2 to the " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in a 177 Fuel Assembly Plant."

We have not requested nor has Metropolitan Edison provided us witn either gracedures or analyses for cooldown of the reactor coolant system by feed-and-bleed, nor have we perforned such evaluations. We, tnerefore, do not know wnether feed-and-bleed can be utili:ed to achieve cold shutcown. However, sufficient water is availacle in tne borated water storage tank (SWST) for at least 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> of feed-and-bleed oceraticn with wo HPI oumes. Af ter the

, SWST has been emptied feed-and-bleed could be continued for an indefinite period by reinjection of the water " bled" from 'he system and stored in the containment sumo. A primary objective of the operator througnout this time would be to reestablish either main or emergency feedwater flow to the steam generators.

The majority of the components of these systems are located outside containment and would be available for service. Once feedwater flow was estabit:hed, one primary system would be cooled and decressurized utili:ing the steam generators.

See the NRC response to UCS Centention 1 question 17 for a discussion of experience using feed-and-bleed.

Question 69 If there is a loss of steam in the secondary system which results in failure of the turbine-driven feedwater pumps, will both motor-driven pumps be required to supply the requisite amount of feedwater? Does this meet tne usual single-failure criteria since it appears tnat a redun-dant system requires multiple comoonents to operate?

(Witness Jensen)

Response: One EF'i pump can supply adequate feedwater for decay heat removal for all postulated accidents and transients. Therefore, the single failure cri erion is satisfied. A single motor-driven emergency feedwater pump would t

have the capability to deliver 460 gpm to the steam generators. As discussed on page Cl-4 of NUTEG-0680, the NRC evaluated the effect of supplying 500 gpm af emargency feedwater following a loss of main feedwater. The conclusion of this evaluation was that although the PORV might be opened by high reactor coolant pressure, the amount of coolant loss would be bounded by the small-break LOCA analyses. The LCCA analysis perfomed for a stuck-open unisolated PORV indicated that no core uncovery or damage wou' d occur if only 300 gpm

.a.

of emergency feedwater were available. This analysis is cescribed in the B&W report " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant, May 7,1979, Section 5.2.3" and Supple-ment 3 to this document dated May 24, 1979. The coolant lost from a PORV cycling on high pressure would be less than that from a stuck open PORV.

Since the stuck open PORV LCCA analysis would bcund a loss of feedwater event witn 460 gpm of emergency feecwater, we concluce that the flow of one motor-driven emergency feedwater pump would ce adequate to cool tne reactor core.

Question 6h. Can the turoine-driven pumos and valves be coerated on Direct Current, or are they dependent upon the Alternating Current safety buses?

('4f tness Wer niel)

Response: The TMI-1 turhine-driven EF4 train consists of one turoine-driven pumo and its associated flow path (including valves). This train can operate to supply feedwater on direct current power sources only as described in the TMI-l Restart SER, NUREG-0680, page Cl-9 and 10.

Question 51. Wii' the reliability of the emergency feedwater system be greatly improved upon conversicn to safety-grad), and is it the licensee's and staff's position that t.N improvement is enough such that the feed-and-bleed back-up is not required?

(Witness Wer niel)

R,esponse: Based on knowledge of the imorovement in reliability gained by eliminating first order failure sources, it is tne staff'- judgement tnat the reliability of the emergency feedwater system will be imoroved once tne fully safety-grade system is installed. The single failure proolem associated

_g-with the integrated control system /non-nuclear instrumentation described in the resconse to 6a and b above will be eliminated.

In addition, various other hardware, proce-dural and administrative imorovements as identified in the TMI-l Rescart SER, NUREG-C680 under Order Item la. should enhance emergency feedwater system reliabili ty.

However, a quantitative reassessment of the reliability of the fully safety-grade EF4 system nas not been perfomed. The feed-and-bleed cack-up is not required by the staff and, derefore, need not meet all recuire-ments of a safety system. However, it is recognized as additional defense in depth for providing core cooling in the very unlikely event that emergency feecwater is lost, and is, therefore, required to be available.

Question 6j. Will the short-term acticis proposed improve the reliability of the emergency feedwater system to the point where restart can be pemitted?

('dt tness Wemiel)

Response: Yes. The proposed short-tem modifications as described in the TMr-1 Restart SER will improve emergency feecwater system reliacility to the point where restart can be pemitted. This is discussed in detail in the TMI-l Restart SER, NUREG-0680, p.,ge C8-37.

Question 6k. Question 6 should be addressed with reference to Florida Power i Lignt Co. (St. Lucie, Unit 2), ALA8-603, (July 30,1980); i.e. wnether loss of emergency feecwater is a design basis event notwithstanding whether design criteria are met.

(Wimess Wer-niel)

Resconse: A loss of emergency feedwater (caricurrent with an accident or trarsient) is not a design basis event for the #cilowing reascns:

. (a) In ALAB-603, the Board refers to Standard Review Plan Section 2.2.3 as the criterton for acceptabfitty of the plant design to mitigate the assumed event. This criterion was intended by the staff to be applied only to external plant ha:ards such as nearty transportation of toxic.

gases or explosives, and not events within the plant. It is, therefore, not appitcable to a costulated loss of emergency feedwater.

(b) The staff perforns a detenninistic review of the emergency feeowater system to verify ccmoliance with applicable General Cesign Criteria.

(c) Addf tfonally, we have applied reliability techniques as a tool for improving emergency feecwater (EFW) system reliability. The results of this evaluation were used to identify and correct te primary sources of system unreliability. Requirements were deternined based on these rei f abili ty studies. This effort has been included in our review of TMI-l for restart, and is discussed in the TMI-l Restart SER, NUREG-0630.

(d) THE NRC has not estabitsbed a numerical safety goal at this time. Work in this area is proceeding as described in SECY-80-379, "Procosed Plan for Ceveloping a Safety Goal," dated August 12, 1980. Until guidance is established, reliability studies will continue to be used as indicated in (c) above.

(e) The single failure crf terion continues to be emoloyed for design basis events as indicated in SECY-77-439, " Single Failure Criterion," dated August 17, 1977. Additionally, probabfifstic methods will be used as a tool for furter insight such as described in (c) acave.

. (f) Based on the EFR system design and the modiffcations to be implemented as de rcribed in the TMI-l Restart SER, NUREG-0680, we believe that further additional hardware changes will not significantly imorove EFW reliability.

The ommon cause failure mode as a result of operator error still remains as the dominant source of system unreitability. This failure mode is be no further minimi7.ed with improvements in the human factors aspects of the plant, i.e., imoroved operating and emergency procedures, improve-ments in instrumentation, and continuous operator trailing.

Jared S. Wermiel Professional Qualifications Auxiliary Systems Branch Division of Systems Integration Office of Nuclear Reactor Regulation I am a Reactor Engineer in the Auxiliary Systems Branch in the Division of Systems Safety. Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission.

In this position I perform technical reviews, analyses, and evaluations of reactor plant features pursuznt to the con-struction and operation of reactors.

I' received a Bachelor of Science Degree in Cnemical Engineering from Drexel University in 1972. Since 1972 I have taken courses on PWR and BWR System Operation, Reactor Safety, and Fire Protection.

My experience includes seven years with the Bechtel Power Corporation as a Systems Design Engineer engaged in the design of various nuclear power plant auxiliary and balance of plant systems. These have in-cluded cooling water systems, water treatment systa.cs and fire protec-tion systems.

I joined the Auxiliary Systems Branch of the Comission in March, 1978.

Since joining the Commission I have performed safety evaluations on nuclear pcwer plant auxiliary systems including auxiliary feedwater systems for the Virgil C. Summer Nuclear Station, Palo Verde Nuclear Generating Station, Waterford Steam Electric Station, Diaolo Canyon Nuclear oower Plant, 3yron/Braicwood Stations and Trojan Nuclear Plant.

I have also reviewed various tooical reports and provided comments on crocosed ANSI Standarcs dealing with various ~.uxiliary systems.

. I have responsibility for the review of the folicwing nuclear power plant auxiiiary systems and concerns: new and spent fuel storage, spent fuel pool cooling, fuel handling, service water, component cooling water, condensate storage, ultimate heat sink, instrument air, chemical and volume control, main steam isolation valve leakage control, heating ventilating and air conditioning, portions of tne main steam system, main feedwater, auxiliary feedwater, hign and moderate energy pipe breaks, f' cod protection and inter-nally generated missiles.

I am a registered Professional Engineer in the State of Maryland.

I am an Associate Memoer of the American Institute of Chemical Engineers.

'#ALICN L. JENSEN, JR.

PRCFESSICNAL QUALIFICATIONS I m a 5anier Nuclear Engineer in me Reactor Systams 3rancn of me Nuclear Requiatery C:cmission. In this position I am rescensible for ce technicai analysis anc avaluation of the puslic health and safety as ects of reac se-systems.

Fr:m June 1973 to Cecameer 1979, I was assigned to the Sulletins anc Greers Task Force of the Nuclear Requiatory C mmission.

I ;ar.icipatac in the Ort:aration of NUREG-0565, " Generic Evaluation of Small 3rtak L:ss-of-Coolant Ac:ican. !anavier in Saccock & 'diic:x Designac 177-FA Operating Plants."

Prem 1972 to 1975, I was assigned to the Containment Systams 3rsnch of me

• RC,' AEC, and f em 19,~ a 1979, I was assigned ts m e Analysis 3 ranch of tae NRC.

In these ;csitions I was responsible for the devoic:=ent, anc evaluation of c:=utar progrus and techniques to calculate the riac* r systam and

ntaf =ent systam tspense.s ;cstuTated loss-of-::oiant ac..dcents.

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• tenivec an M.S. cegree in Nucitar Ingineering at the "atact ic University :f Amarica in *.353 anc a 3.5. cegree in Nuclear Engineering at ftississi:ci State University in '363.

I am a gricuata Of the Cat Ridge Sencol for Reac se Tecanclegy, 1963-1964 I am a cam:ar f the American Nuclear Society.

I am the autner Of three scientific pacers csaling with the r:scense Of 3&W L:ss f-Coolant Ac:icents and have au'herac cne scientific pacer reactors.

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EDWARD LANTZ DIVISION OF SAFETY TECHNOLOGY U.S. NUCLEAR REGULATORY COMMISSICN PROFESSICNAL QUALIFICATIONS As an Engineering Systems Analyst in the Operating Experience Evaluation Branch I am responsible for reviewing and evaluating experience related to the safety of nuclear power plant operation.

I have a Bachelor of Science degree in Engineering Physics from the Case Institute of Technology and a Masters of Science degree in Physics from Union College and a total of 30 years of professional experience, with over 20 years in the nuclear fiel d. My experience includes work on reactor transients and safeguards analysis, nuclear reactor analysis and design, rescarch and development on nuclear reactor and reactor control concepts and investigations of their operational and safety aspects.

I have held my present position with the Commission since May 1,1980. My previous position, which I held for about four and one half years, was Engineering Systems Analyst in the Plant Systems Branch where I was resconsible for technical reviews and evaluations of component and system designs and operating characteristics of licensed nuclear power reactors. Prior to that I was a Project Manager in the Gas Cooled Reactors 3 ranch, Division of Reactor Licensing, U.S. Nuclear Regulatory Ccemission, where I was responsible for the technical review, analysis, and evaluation of the nuclear safety aspects of acplications for construction and operation of nuclear power plants.

For about ten years prior to that I was Head of the Nuclear Reactor Section in NASA. My section was responsible for the development and verification of nuclear reactor analysis computer crograms, conceptual design engineering, and development engineering contracting. Prior to my employment with NASA, I was a nuclear engineer at the Knolls Atomic Power Labc-atory for about six years, where I worked on the safeguards and nuclear desi;n of the 53G reactors and the initial development of the nuclear design of t..c 55G reactors. Previous excerience includes system engineering and electrical engineering with the General Electric Company and electronic development engineering with the Victoreen Instrument Ccmpany.

8ebcock&Wilcox s e,ce,e,m,c,cc, P.O. Sex 1250. :.ync. :u. g vs. 245C5 7eis:nene::SC41234 5111 May 12, 1979 0:. 1. J. Ma:: son, Directo:

Divisien of Systes safety Office of Nuclear Reac:=: Regulatica U.S. Nucles: Reg..la:Ory Cc~ issica Washington, D.C.

20535

Reference:

"Ivaluatien of Transien: 3ehavior and Snail Reactor C clan:

System 3:eaks in the 177 Fuel Assenbly Plant," May 7, 1979.

Oear Jr. Mattsen:

he Staff's review of the reference has resul:ed in an NRC recuest for addi:icnal inferna:1ca en See:1:n 6 "S-nil 3:eak Analyses." Specifi-cally, the 5:aff recuested the folleving analyses:

"Snall 3:eak in the Pressuri:e: (?CR7) With No Auxiliary Teedva e: and One

'3' **p" 35.* has perferned the requested analysis and ::: luded :ha: :he systen is nain:sined in a safa c:ndi:icn vich the core ra-ming c:vered out

o a: leas: 30 ninutes. This calculatica involves :he conservative in-pu: assu=p:1cn of 1.2 x ANS decay heat.

• he results =f the analysis are a:: ached :o :his le::e and have been identified as Supplanen: 1 to See:1:n 6 of :he :sference.

Te also have ce=ple:ed a s nilar calculation using 1.0 : ANS decay heat.

~his analysis was carried cut to app xi=stely 4700 seconds where :he pressure tu: ed 2::und.

~~ae core remains : vered th::ughcut the :::n-sian:.

• he resul:s of the analysis are attached to this le::e and are designated as Supplanen: 2 := S4.c:1:n 5 of :he reference.

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INFORMATION REPORT

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a licensing "eView CCol EC assure.*el'able systers 038 13 as ene ele:ent of the defense in depth a;; reach to.eacter sa.'e ty.

~"'.e Reacter Safe.y Study indi:ates taa:. its use had led to a generally acceptable level Of t,a.-tware r*du".dancy in OOst syste.s 1. pC.~ta".0 OC safety. 500e

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It is expected Snat pr:Catilistic metheds of the type used 1.n the Reacter Safety Study will gracually ecce into increasing use and su;;1e ent

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