ML19338F369

Testimony Re ASLB Question 6 Re Emergency Feedwater Sys Reliability.Emergency Feedwater Following Main Feedwater Transient Is Not Type of Accident Requiring safety-related Equipment Protection.Prof Qualifications Encl

ML19338F369
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Site:	Crane
Issue date:	10/14/1980
From:	Jensen W, Lantz E, Wermiel J Office of Nuclear Reactor Regulation
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE-ATOMIC SAFETY AND LICENSING BOARD 4

In the Matter of

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METROPOLITAN EDISON COMPANY, Docket No. 50-289 et. al.

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(Three Mile Island Nuclear

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. Station, Unit 1)

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NRC STAFF TESTIMONY OF:J. WERMIEL~, W. JENSEN E. LANTZ, AND B. 80GER REGARDING EMERGENCY FEEDWATER SYSTEM RELIABILITY (BOARD QUESTION 6)

~(Tr.2394-96)

Emergency Feedwater Reliability Question 6a.

Is a loss of emergency feedwater following a main feedwater transient an accident which must be protected ~against with safety-grace equipment? Would such an-accident' be caused or aggravated by a loss of non -

nuclear instrumentation, such as-occurred at Oconee?

(Witness Wermtel)*

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

With respect.to the. 0conee Incident, a complete loss of all feedwater will not result from a failure in the integrated control system /non.1uclear instrumenta-

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

• Note:. Name in plarenthesis indicates. preparer of response.

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in the iMI-li Restart SER, NUREG-0680, pages C8-35,: 36 and 37) that will completely.

eliminate.any:intertie between the ICS/NNI and ERf systems. These modifica-tions will upgrade the.ER4 system to a fully safety-grade system.

Prior to -

. installation of-the fully safety-grade system, an Oconee. type event may result-in a.momentry loss 'of ER4. However, this' situation.would be detected by the operator-tarough the ERJ flow indicators and steam generator level indication which are separate > from ;the NNI-as described in the TMI-l Restart SER, NUREG-

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0680. The operator; can then take the necessary manual action in the control room to open tne ERf control ~ valves and restore feed flow.

(WitnessJensen)

. In the unlikely event that both main and emergency feedwater cannot be

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restored, the operator at TMI is instructed to actuate.High Pressure Injec-tion:as described in Inadequate Core Cooling Emergency Procedure EP1202-39.

This action.will' provide adequate cooling to the core by feed and bleed if two high_ pressure injection pumps'are ave.ilable. The high pressure injection system operates independent of NNI/ICS.

Question;6b.

In what respect is the emergency feedwater system vulnerable L-l l

to non-safety-grade syste failures and to. operator errors?

(Witness Wermiel) l L:

Response: : The' emergency feedwater system currently meets all requirements l-

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egrade' 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 flow co'ntrol valves. - Implementation of the safety-grade modification j

' to the automatic initiation' design, which will eliminate the single failure.

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- possibility, wf111 correct -this deficiency.

In addition,'a question has been raised by the licersee~in Licensee Event Report 80-012/0IT-0 dated July 11, 1980, concerning the adequacy of the environmental' qualf #ication for the emergency feedwater system components in the event of a main steam line break in1the Intermediate Building. This area-is being reviewed in connection with IE1 Bulletin 79-018.

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. With regard to the. vulnerability of the EFW system to operator errors, the EFN. system is' vulnerable to operat r errors as is any safety-grade system.

For example, the operator may leave alves closed or turn off pumps. However, improved emergency:and operating procedures coupled with operator training in.

these procedures (as described in the TMI-1 Restart SER, NUREG-0680) 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 Lantz)

Response

We have reviewed tha. Licensee Event Reports for plants with

- safety-grade emergency feedwater (EFW) systems. The available data for plants j

i 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

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occurred.did lnotidefeat the -functional capability of the system.

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'G on EP4 system success ontdemand is not maintained. However, it should be-

noted that:all plants perform routine periodic EPA system surveillance testing

..in'accordance with specific plant Technical Specifications. Except for the

.following reported cases of common cause and operator-induced failure, which -

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resulted in an ove'rall system failure, the system was found to be operable when' initiated. These are cases where sufficient emergency feedwater was

.not available,.although emergency feedwater was not required at the time to cool-the reactor.

Plant Date LER#

Description

.Ginna.

12/14/73 73-01T The' suction was lost on two emergency-feedwater pumps during

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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

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third.feedwater pump started i

but tripped because of foreign ~

matter in the governor.

L Kewaunee 1 11/05/75 75-01T During startup operations resin

' beads clogged the strainer to

.all emergency feedwater pumps.

Haddam Neck 07/05/76 76-03L When the plant was'in the startup mode, both emergency'.feedwater pumps were vapor bound due to.

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a leaky check valve.'

It should~ be noted that prior to commercial operation at the Davis Besse.

. plant, there was. one event lin which the safety. grade emergency feedwater l

.: system failed on demand.

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-~5' Question 6d.: ' What operator action-is required to operate in a feed-and-bleed mode following a loss of emergency feedwater?

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'(Witness Boger)

Response: 'According to EP. 1202-26A', " Loss of Steam Generator Feedwater to Both

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' 0TSGs," if both main and emergency feedwater are lost, the operator is directed to initiate' High Pressure Injection (HPI). HPI is actuated by depressing two pushbuttons on the main control. board.

The next operator action is to verify L

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 PORV is cycling to maintain RCS pressure at approximately 2450 psig. At this

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point-the operator has provided a makeup supply to the RCS (feed) and a relief-path to remove core heat (bleed).

If the PORY or PORV block valve fail to respond (open), the RCS ' safety valves will relieve at 2E00 psig to provide the

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t bleed path.

Subsequent operator actions include. throttling the HPI discharge valves 'to maintain at-least' a 50*F margin of subcooling and attempting to restore a l

.feedwater supply to the steam generators from the main or emergency feedwater system or the condensate ' system.

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. Question-6e..If-the emergency feedwater. system were to fail, what assurance

.do;we have that thel 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

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- under. two-phase mixtures.

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(Witness Jensen)

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

Question 6f. Can the system be taken to cold shutdown with the feed-and-bleed cooling only?. Are both high' pressure injection (HPI) pumps 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 there been

.any e.xperience using that system?

(Witness -Jensen)

Response: Analyses by Babcock.and Wilcox indicate two high pressure injection pumps are required to adequately cool the core by feed-and-bleed for the first three hours if decay heat is calculated using 1.2 times the ANS-5 decay

-heat model as required for LOCA an'alysis in Appendix K to 10 CFR 50. After three hours only one pump would be' required.

If a best estimate decay heat model is utilized (1,0 times the ANS-5 decay heat model) the analyses indicate that only one HPI pump would be required. See letter from J. Taylor, B&W, to R..'lattson, NRC, May 12, 1979, which transmits Volume 1 Section 6.0 - Supplements 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 with either procedures o'r analyses for cooldown of the reactor coolant system by feed-and-bleed,.nor.' have wr perfonned such evaluatio..s. We, therefore, da not know whether feed-and-bleed can be utilized to achieve cold shutdown. However, sufficient water is available in the borated water storage tank (BWST) 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 operation with two HPI pumps. After the

BWST has been emptied feed-and-bleed 'could be continued for an indefinite period' by reinjection of the water " bled" from the system and stored in the containment sump. A ' primary objective of the. operator throughout 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 established, the primary system would be cooled and depressurized utilizing '.he steam generators.

See the NRC response to UCS Contention 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 that a redun-dant system requires multiple conponents to operate?

(WitnessJensen)

Response: One EFW pump can supply adequate feedwater for decay heat removal for all postulated ' accidents and transients. Therefore, the single failure criterion is satisfied. A single motor-driven-emergency feedwater pump would lh' ave. the capability. to deliver 460 gpm to the steam. generators. -~ As discussed

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on page Cl-4 of NUREG-0680, the NRC evaluated the effect of supplying 500 gpm of emergency feedwater following a loss.of main feedwater. The conclusion of this evaluetion 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 LOCA-analysis performed for a stuck-open unisolated

.P0RV indicated that-no core uncovery or damage would occur if only 300 gpm 4

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of emergency -feedwater were available. This analysis is described 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 6.2.3" and Supple-ment 3 to this document dated May 24,1979. _ The coolant lost from a PORY cycling on high pressure would be less than that from a stuck open PORV.

Since the stuck open PORV LOCA analyst.s would bound a loss of feedwater event with 460 gpm of emergency feedwater, we conclude _ that the flow of one motor-driven emergency feedwater pump would be adequate to cool the reactor core.

Question 6h-Can tne turbine-driven pumps and valves be operated on Direct Current, or are they dependent upon the Alternating Current safety buses?

(Witness Wrmiel)

Response: The TMI-1 turbine-driven EP4 train consists of one turbine-driven pump and its associated flow path (including valves). This train can operate to supply feedwater on direct current power sources only as described in the-

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TMI-1_ Restart SER, NUREG-0680, page Cl-9 and 10.

Question 61. Will the reliability of the emergency feedwater system be greatly improved upon conversion to safety-grade, and is it the licensee's

~ and staff's position' that the improvement is enough such that the feed-and-bleed back-up is not required?

.(WitnessWermfel)-

Response

Based on knowledge of the improvement in reliability gained by eliminating first order failure sources, it is the staff's judgement that the

' reliability of the emergency feedwater' system will' be improved once the fully safety-grade' system is installed.

The single failure problem associated j

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

In addition, various other hardware, proce-dural and administrative improvements'as identified in the T!11-1 Restart SER, NUREG-0680 under Order Item la. should enhance emergency feedwater system reliability. 'However,.a quantitative reassessment of the reliability of the 4

fully safety-grade EFW system has not been performed. The feed-and-bleed back-up is not' required by the staff and, therefore, need not meet all require-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 feedwater is lost, and is, therefore, required to be available.

Question 6j. Will the short-term actions proposed improve the reif ability of the emergency feedwater system to the point where restart can be permitted?

(Witness Wermiel)-

Response: Yes. The proposed short-term modifications as described in the

- TMr-1 Restart SER will improve' emergency feedwater system reliability to the point where restart can be permitted. This is discussed in detail in the TMI-l Restart SER, NUREG-0680, page C8-37.

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Question 6k. ~ Question 6 should be ' addressed with reference to Florida Power &

Light Co. -(St. Lucie, Unit 2), ALAB-603 (July 30,1980); i.e. whether loss

- 'of emergency feedwater is a design basis event notwithstanding,whether design icriteria are met.

(Witness Wermiel)

Response

A' loss of emergency feedwater (concurrent with an accident or transient) is' not a design basis event for the following reasons:

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'(a)? iIn ALAB-603, the' Board refers to Standard Review Plan Section 2.2.3 as the criterion.for acceptability of the' plant design to mitigate the 9

assumed' event. This criterion was' intended by the staff to be applied

-only~to; external plant hazards such as nearby transportation of toxic

_ gases o'r explosives, and not events within the plant.' It-is, therefore, not applicable to a postulated loss of emergency feedwater.

(b) The staff perfonns:a deterministic review of the emergency feedwater system to verify compliance with applicable' General Design Criteria.

(c) Additionally,' we have applied reliability techniques as a tool for improving. emergency feedwater (EFM) system reliability. The results of this evaluation were used to identify and correct the primary sources of system unreliability. Requirements were determined based on these reliability studies. -This' effort has been included in our review of

-TMI-1 for restart, and is' discussed in the TMI-1 Restart SER, NUREG-0680.

-(d). THE.NRC has not established a numerical safety goal at this-time. Work ifn this area is proceeding:as described in SECY-80-379, " Proposed Plan for Developing a-Safety Goal,"' dated August 12, 1980. Until guidance is established,: reliability _ studies will continue to be used

asiindicated in (c) above.

- ;(e) The single' failure: criterion continues' to be. employed for design basis

. events,as) indicated tin' SECY-77-439,i" Single Failure; Criterion," dated fAugust 17,f 1977;:(Additionally,1probabilistic methods will 'be used as a toolf for further insight such asjdescribed in (c) above.

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(f) Based onEthe.ERI. system design and the modifica'tions to'belimplemented as' described in: the THI-1, Restart SER NUREG-0680, we believe that further additional hardware changes will;not significantly improve EFW reliability.

The common cause failure mode as a-result of Operator error still remains as the' dominant source of system unreliability. This failure mode is being further minimized with. improvements l'n the human factors aspects of the plant,1.e., improved ~ operating and emergency procedures, improve-

- ments in _ instrumentation, and continuous operator training.

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1 lJared S. Wemiel Professional Qua'lifications

' Auxiliary Systems Branch Division of Systems Integration Office.of, Nuclear Reactor. Regulation 1

1 am a Reactor Engineer in the Auxiliary System's Branch in the Division of Systems Safety, Office of Nuclear Reactor Regulation, U.S. Nuclear

. Regulatory Comission.

In this position I perform technical reviews, analyses, and evaluations of. reactor plant features pursuant to the con-structionandoperationofreactors.

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I' received a Bachelor of Science' Degree in Chemical Engineering from Drexe1 D iversity in 1972. Since 1972 I have taken courses on PWR and

.BWR System Optration, Reactor Safety,'and Fire Protection.

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My experience includes seven years wich the-Bechtel Power Corporation y

as a Systems Design Engineer enosged in the design of various nuclear power plant auxilir.ry and balance of plant systems. These have in-cluded cooling water systems, water treatment systems and fire protec-tion systems.

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.I joined the ' Auxiliary Systems Branch of the Com;ssion in March, 1978.

Since joining-the Comission.I have perfonned safety evaluations on nuclear. power plant auxiliary systems including auxiliary feedwater systems for the Virgil.C. Sumer Nuclear-Station, Palo Verde Nuclear Generating Station.. Waterford Steam Electr*c Station, Diablo Canyon Nuclear Power Plant, Byron /Braidwood Stations and Trojan Nuclear Plant.

(I have also reviewed various~ topical. reports ed provided comments on 3

. proposed ANSI l Standards dealing with various auxiliary systems.

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I have responsibility for the review of the following nuclear power plant auxiliary systems and concerns: new 'and spent fuel storage, spent fuel pool

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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 the main steam system, main 'tedwater, auxiliary feedwater, high and moderate energy pipe breaks, flood protection and inter-nally generated missiles.

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~1 am a registered Professional Engineer in the State of Maryland.

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

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WALTON L. JENSEN, JR.

PROFESSIONAL QUALIFICATIONS i

I am a Senior Nuclear Engineer in the Reactor Systems Branch of the Nuclear Regulatory Commission. In this position I as responsible for the technical analysis and evaluation of the public hea1th and safety aspects of reactor-

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systems.

From June 1979 to December 1979, I was assigned to the Bulletins and Orders

- Task Force of the Nuclear Regulatory Commission.

I participated in the preparation of NUREG-0565, " Generic Evaluation of Small Break Loss-of-Coolant

-A::ident Sahavior in Babcock & Wilcox Designed 177-FA Operating Plants."

From 1972 to 1976, I was assigned to the Containment Systems Branch of the NRC/AEC, and from 1976 to 1979, I was assigned to the Analysis Branch of the NRC.

In these positions I was responsible for the development,and evaluation of computer programs and techniques to calculate the reactor system and '

containment system response to postulated loss-of-coolant accidents.

From 1967 to 1972, I was employed by the Babcock and Wilcox Company at Lynchburg, Virginia. There I was lead engineer for the development of less-of-coolant computer programs and the qualification of these programs by comparison with experimental data. -

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4 D""D '"D NY Ed a b-d, From 1963 to 1967, I was employed by the Atomic Energy commission in the Division of_ Reactor Licensing.

I assisted in the safety reviews of large power reactors, and Inled the reviews of several small research reactors.

I receivec an M.S. degree in Nuclear Engineering at the Catholic University of i

America in 1958 anc a B.S. degree in Nuclear Engineering at liississippi. State University in 1953.

I am a graduate of the Oak Ridge School for Reactor Technology, 1963-1964 I am a macter of the American Nuclear Society.

I am the author of three scientific papers dealing with the response of S&W reactors to Loss-of-Coolant Accidents and have authored one scientific parer daaling with containment analysis.

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' PROFESSIONAL QUALIFTCATICNS LIST BRLE A. BOGER F4acatien-June 1971 Received BS!"E - Uriversity of V6girla June 1972 Received PESE - Crdversity of Virginia W k F.rerience -

June 1972 to Virginia Electric and Power Cc=nany

.h=e 1977 Surry !belea: Power Station Assista.: Engineer - Perferred starr.:p testing en thi:

No. 2..

E. ghee:

Assisted the Sgereisc -Egineering Se: vices;

ained for and received a.Senic: Reae:c: Cpera::: License.

Supervisor - Engineering Services - Directed the activities of the ensite engineering staff.

June 1977 to Virginia Electric and Power Coccany Septec6e: 1977 Rich end, Virginia Supervisor - Ibelear Engineering Services - Directed the activi:ies of the effsite engineeri.4 staff in st=per: ef j

Surry Power Station.

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Oc:cber LC77 to U. S. ?belear Regulatory Ccccission

' Presen:

Bethesda,-Maryland Reactor Engineer in the Ooerater Licensing Branch - A%-

ister licensing e.u._' nations to nuclea pcr.e: plant and research reactor personnel.

- h efessic.a1 Affiliatiens Regis ered ?:cfessional Engineer - State of Virgi.-la PW '-' American Ibeiear Society a

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&S Participation in M Activities Bruce A. Bege

. Nove=ber 1978,' April 1980: A6:inistered operate: license exednations en Unit

One.

Neve-ber 1978, March 1979, March 1980: Ai:inistered cpera:c: license exa:-l.utiens en thi: I k.-

March - April 1979: Me=ber of the M-2 energency response tea =, assisted in the preparation of ecargency and 'centingency procedres.

.7uly 1979 - P esent: Mecber of the M. Technical Sucpc:: S:aff, cend.:cted audi:

exadnations en pcs:-acciden: installed equipment en M-2. %1sc partici-pared in the review of training and procedres in conjunction 'th the'M-1 zestart p cgrams. ?J.s included prepara:icn of SER inpu:s and :es:Lm...

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EDWARD LANTZ DIVISION OF SAFETY TECHNOLOGY U.S. NUCLEAR REGULATORY COMMISSION PROFESSIONAL 100ALIFICATIONS As an. Engineering Systems 'AnalystLin.the Operating Experience Evaluation Branch I am responsible for reviewing and evaluating.e,xperience related to the safety J 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? rom Union College and.

f a total of 30 years of' professional-experience, with over 20 years in1the nuclear-field.. My experience includes; work on reactor. transients and safeguards analysis, nuclear reactor analysis'and design, research and development on nuclear reactor i

and reactor control concepts and investigations of their operational and safety aspects.

Ihave~ held'mypresentpositionwiththe[ Commission-sinceMay' 1,1980.1MypreYious position, which I_ held for abcut four and one half years, was Enginee' ring Systems Analyst in.the Plant' Systems' Branch where I was responsible 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 Branch, Division of Reactor Licensing, _ U.S. Nuclear

. Regulatory. Commission, where ~I was responsible for the technical review, analysis, and evaluation of the nuclear safety aspects of applications for construction and operation of nuclear; power plants.

For about ten years prior to that I was Head of the NuclaarLReactor Section -in NASA. My section was responsible for the

- ' development'and verification off nuclear reactor analysis computer programs

. conceptual' design engineering, and development engineering contracting.

Prior

-to my employment with NASA," I was a nuclear engineer at the Knolls Atomic power Laboratory for about six years, where I worked on:the safeguards and nuclear design of the S3G reactors and the initial ~ development of _the nuclear design. of j

the S5G reactors. ' Previous experience includes system engineering and. electrical-

- engineering with the General Electric Company and electronic'de,velopment engineering with.the Victoreen-Instrument Company.

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