ML20104B856

Forwards Description of Plan for Resolution of TMI Action Item II.K.3.5 Re Automatic Trip of Reactor Coolant Pumps,Per Generic Ltr 83-10c

ML20104B856
Person / Time
Site:	Seabrook
Issue date:	04/08/1983
From:	Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE, YANKEE ATOMIC ELECTRIC CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
Shared Package
ML20104B828	List: ML20104B823 ML20104B856
References
TASK-2.K.3.05, TASK-TM GL-83-10C, SBN-498, NUDOCS 8502040283
Download: ML20104B856 (11)

Text

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Attachment 3 ,

' :, Page 1 of 6 l SEAstOOK STAT 10N i b Engineesbis Ogless l 14y1 Wenesser Reed  !

Pd$c Service of New HampsNro g ,

bec: J.P. Cady J.E. Tribble UE&C&W(SB-15531)

A.C,.Cerne.,,k D.f,Floyd i AS G i

April 8,1983

. J.H. Herrin R.J. Harrison i G.F. Mcdonald L. Walsh SBN-498

- D.N. Merrill A.E. Ladieu T.F. 54.2.99 D.E. Moody P. Swanson i

NRC Chrono P.L. Anderson R.P. Pizzuti W.N. Fadden

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. souderos United ' .tes Nuclear Regulatory Commission

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Washingtc,n, D. C. 20555 Ropes & Gray (Dignan/Ritsher/ Gad) l A.M. Shepard Attention

Mr. Darrell G. Eisenhut, Director Division of Licensing B 99

References:

(a) Construction Permits CPPR-135 and CfPka13R"!Mcket Nos. 50-443 and 50-444 (b) USNRC Letter, dated February 8, 1983, " Resolution of M I Action Itea II.K.3.5, ' Automatic. Trip of Reactor Coolant Pumps, ' (Generic Letter No.83-10c)," D. G. Eisenhut to All Applicants with Westinghouse (W) Designed Nuclear Steam Supply Systems (NSSS)

Subject:

Response to NRC Generic Letter No.83-10e

Dear Sir:

We have enclosed a detailed description of our plan for resolution of TMI Action Item II.K.3.5 in response to Generic Letter 83-10c [ Reference (b)].

As is evident in the enclosed response, we are participants in the Westinghouse Owners Group effort to resolve this ites and therefore, our schedules for plant specific submittals are tied to schedules developed by the Owners Group.

Very truly yours, YANKEE ATOMIC ELEC*RIC COMPANY 8502040283 850128 PDR ADOCK 05000443 E PDR J. DeVincentis Project Manager .

ALL/fsf ec: Louie Wheeler, Project Manager Division of Licensing ,

Atomic Safety and Licensing Board Service List 1000ElmSt..P.O. Box 330.Monchester.NH03105 TelerAone(603)669-4000 TWX7102207595

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Attachment 3 PLAN FOR RESOLUTION OF TMI ACTION ITEM II.K.3.5 Page 2 of 6

" AUTOMATIC TRIP OF REACTOR COOLANT PUMPS" INTRODUCTION The criteria for resolution of TMI Action Plan Item II.K.3.5, " Automatic Trip of Reactor Coolant Pumps" were stated in letters from Mr. Darrell G. Eisenhut  ;

of the Nuclear Regulatory Commission to all Applicants and Licensees with j Usstinghouse designed Nuclear Steam Supply Systems (83-10 e and d), dated  :

PebrGary 8, 1983. The following represents the plan for demonstrating 1 compliance with those criteria. In order to avoid confusion, the overall  !

philosophy and plan will first be stated. Then, each section of the  :

attachment to NRC Letters 83-10 e and d will be addressed as to how the j l

overall plan responds to each NRC criteria.  ;

OVERALL PLAN I l

i In the four years that have passed since the event at three Mile Island, f Westinghouse and the Westinghouse Owners Group have held steadfastly to several positions relative to post-accident Reactor Coolant Pump (RCP) ,

operation. First, there are small break LOCAs for which delayed RCP trip can I result in higher fuel cladding temperatures and a greater extent of j sircalloy-water reaction. Using the conservative evaluation model, analyses f for these LOCAs result in a violation of the Emergency NCore Cooling currently System approved  !

' _(ECCS) Acceptance Criteria as stated in 10CPR50.46. [

Westinghouse Evaluation Model for small break IACAs was used to perform these  ;

analyses and found acceptable for use by the NRC in Letters 83-10 e and d. l hrefore, to be consistent with the conservative analyses performed, the RCPs  ;

should be tripped if indications of a small break LOCA exist.

Secondly, Westinghouse and the Westinghouse Owners Group have always felt that the RCPs should remain operational for non-LOCA transients and accidents This where their operation is beneficial to accident mitigation and recovery.

position was taken even though a design basis for the plant is a loss of off-site power. Plant safety is demonstrated in the Pinal Safety Analysis Reports for all plants for all transients and accidents using the most conservative assumption for Reactor Coolant Pump operation.

In keeping with these two positions, a low RCS pressure (sympton based) RCP trip criterion was developed that provided an indication to the operator to trip the RCPs for sus 11 break LOCA but would not indicate a need to trip the RCP for the more likely non-IACA transients and accidents where continued RCP operation is desirable. N basis for this criterion is included in the generic Emergency Response Guideline (ERC) Rackground Document (E-0 Easic Reviaton, Appendix A). Relevant information regarding the expected results of seias the RCP trip criterion can be derived from the transients which resulted from the stuck open steam dump valve at North Anna in 1979, the steam generator tube rupture at Prairie Island in 1980, and the steam generator tube Bowever, rupture at Ginna in 1982. The RCPs were tripped in all three cases.

a study of the North Anna and Prairie Island transients indicated that RCP trip would not have been needed based on the application of the IRC trip criterion.

The Ginna event, however, indicated a need to review the basis for the RCP trip criterion to allow continued RCP operation for a steam generator tube rupture for low head SI plants.

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Attachment 3 l ,,

• Page 3 of 6 i Thirdly, it has always been the position of Westinghouse and the Westinghouse

! Owners Group that if there is doubt as to what type of transient or-accident is in progress, the RCPs should be tripped. Again, the plants are designed to mitigate the effects of all transients and accidents, even without RCPWe operation while maintaining a large margin of safety to the public.  ;

existing emergency operating procedures reflect this design approach. ,

Lastly, it remains the position of Westinghouse and the Westinghouse Owners Group that RCP trip can be achieved safely and reliably by the operator when required. An adequate amount of time exists for operator action for the small

break LOCAs of interest. The operators have been trained on the need for RCP 3 trip and the emergency operating procedures give clear instructions on this metter. In fact, one of the initial operator activities is to check if

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indications exist that warrant RCP trip.

e Westinghouse and the Westinghouse Owners Group will undertake a two part program to address the requirements of NRC Letters 83-10 e and d based on the j aforementioned positions for the purpose of providing more uniform RCP trip t criteria and methods of determining those criteria. In the first part of the i program, revised RCP trip criteria will be developed which provides an

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indication to the operator to trip the RCP's for small break 14CAs requiring such action, but will allow continued RCP operation for steam generator tube ruptures, less than or equal to a double-ended tube rupture. The revised RCP

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trip criteria will also be evaluated against other non-LOCA transients and accidents where continued RCP operation is desirable in order to demonstrate that a need to trip the RCPs will not be indicated to the operator for the j more likely cases. Since this study is to be utilized for emergency response

! guideline development, better estimate assumptions will be applied in the i consideration of the more likely scenarios. The first part of the program will be completed and incorporated into Revision 1 of the Emergency Response j

3 Guidelines developed by Westinghouse for the Westinghouse Owners Group. The g

scheduled date for completion of Revision 1 is July 31, 1983.

The second part of the program is intended to provide the . required j- justification for manual RCP trip. This part of the program must necessarily be done after the completion of the first part of the program. The schedule t

j for completion of the second part of the program is the end of 1983.

$ The preferred and safest method of pump operation following a small break LOCA f is to manually trip the RCPs before significant system voiding occurs.

s No attempt will be made in this program to demonstrate the acceptability of j, continued RCP operation during a small break LOCA. Further, no request for an y esseption to 10CFR50.46 will be made to allow continued RCP operation during a ses11 break LOCA.

EgTAILED RESPONSE TO NRC LETTERS 83-10 C AND D H

d Bach of the requirements stated in the attachment to NRC Letters 83-10 e and d j

will now be discussed indicating clearly how they will be addressed. The j organisation of this section of the report parallels the attachment to NRC

] Letters 83-10 e and d.

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Attachment 3 Page 4 of 6 y.e I. Pump Operation Criteria Which Can Result in RCP Trip During Transients and Accidents.

1. Setpoints for RCP Trip The Westinghouse Owners Group response to this section of requirements will be contained in Revision 1 to the Esergency

- Response Guidelines scheduled for July 31, 1983. Seabrook Station will utilize Revision 1 to the Emergency Response Guidelines in the establishment of setpoints for RCP trip. Completion of plant specific proced. ires for Seabrook Station based on the Westinghouse Owners Group Emergency Response Guidelines (including Rev.1) is scheduled for Dec, ember 1983.

a. As stated above, Westinghouse and the Westinghouse owners Group are developing revised RCP trip criteria which will assure that the need to trip the RCPs will be indicated to the operator for LOCAs where RCP trip is considered necessary. The criteria will also ensure continued forced RCS flow for:

1. Steam generator tube rupture (up to the design basis, double-ended tube rupture).

5 2. The other more likely non-LOCA transients where forced circulation is desirable (e.g., steam line breaks equal to or smaller than 1 stuck open FORV).

NorE: Event diag:tosis will not be used. The criteria developed will be symptom based.

The criteria being considered for RCP trip are:

1. RCS vide range pressure y constant I 2. RCS subcooling ( constant i

I 3. Wide range RCS pressure ( function of secondary pressure Instrument uncertainties will be accounted for. Environmental j

uncertainty will be included if appropriate.

l No partial or staggered RCP trip schemes will be considered.

l- Such schemes are unnecessary and increase the requirements for l ,

I training, procedures, and decision making by the operator during transients and accidents.

j b. The RCP trip criteria selected will be such that the operator i

will be instructed to trip the RCPs before voiding occurs at

, the RCP.

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Attachment 3 1

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c. The ctateria developed in Iten la above is not expected to lead to RCP trip for the more likely non-IDCA and SGTR transients.

However, since continued RCP operation cannot be gua'ranteed, the emergency response guidelines provide guidance for the use of alternate methods for depressurization.

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d. The Emergency Response Guidelines contain specific guidance for l detecting, managing, and removing coolant voids that result i from flashing. The symptoms of such a situation are described in these guidelines and in detail in the background document

' for the guidelines. Additionally, explicit guidance for operating the plant with a vaporous void in the reactor vessel head is provided in certain cases where such operation is needed. Seabrook Station will utilize the Emergency Response Guidelines to develop procedures for the detection, management, 1-and removal of Reactor Coolant System voids. Training in the l use of these procedures will be provided.

e. A containment isolation signal (4.3 pois) will result in the l isolation of the RCP seal water return line; however, continued operation of the RCPs is allowable because Primary Component Cooling Water will continue to be provided to the thermal -

barrier heat exchangers, lube oil coolers and motor coolers until the containment spray signal (18 peig) or a low PCCW surge tank level is reached. At this point in the transient, It f

the RCPs would be tripped (if they had not been already). '

should be noted that the containment spray signal does not isolate RCP seal water injection. l

f. Discussed in la and Ic.

2. Guidance for Justification of Manual RCP Trip The Wertinghouse Owners Group response to this section of PSNH requirements will be reported seperately at the end of 1893. i will review the Westinghouse Owners Group guidance for justification '

of manual RCP trip and will provide a plant specific justification for manual RCP trip within three months of receipt of the Westinghouse report.

a. . A significant number of analyses have been performed by Westinghouse for the Westinghouse Owners Group using the currently approved Westinghouse Appendix K Realuation Model for easil break LOCA. This Evaluation Model uses the WFLASH Code.

These analyses demonstrate for ses11 break LOCAs oT concern, if the RCPs are tripped 2 minutes following the onset of reactor l

conditions corresponding to the RCP trip setpoint, the j

predicted transient is nearly identical to those presentedThus, in the Safety Analysis Reports for all Westinghouse plants. '

l the Safety Analysis Reports for all plants demonstrate compliance with requirement 2a. The analyses performed for the Westinghouse Owners Group will be used to demonstrate the f- validity of this approach.

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Attachment 3 '

Page 6 of 6 L$

6 f b. Better estimate analyses will be performed for a limiting j~ Westinghouse designed plant using the _WFLASH computer code with better estimate assumptions. Dese analyses will be,used to determine the minimum time available for operator action for a J

range of break sizes such that the ECCS acceptance criteria of

" 10CFR50.46 are not exceeded. It is expected that the minimum time availab1s for manual RCP trip will exceed the guidance E

- contained in N660. This will justify manual RCP trip for all plants.

j 3. Other Considerations t

a. Information regarding the quality of instrumentation which will be employed to monitor RCP trip setpoint parameters will be i provided to the NRC within three months of the receipt of the Westinghouse report.

b. Seabrook Station will utilise the Emergency Response Guidelines

' to develop procedures for the timely restart of the reactor coolant pumps when conditions which will support safe pump

.startup and operations are established. Training in the use of these procedures will be provided on the Seabrook sita specific simulator.

c. Seabrook Station operators will be knowledgeable / trained in their responsibility for tripping the RCPs when the trip setpoints are reached. De priority of this action and all actions following engineered safety features actuation are also considered.

' II. Pump Operation Criteria Which Will Not Result in RCP Trip During

,l Transient and Accidents.

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The preferred and safest method of operation' following a small break LOCA

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is to manually trip the RCPs. Therefore, there is no need to addrens the

' criteria contained in this section.

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Attachment 4 Page 1 of 5 SB 1 & 2 Ameldsent 52 FSAR - December 1983 overload devices. In addition to the 15 kV switchgear breakers, the medium voltage 15 kV penetrations are also protected by fuses inserted in the feeders out-side containment. These fuses are qualified by experi-ence and seismic testing. The 600 volt system x/R ratio used in specifying the electrical penetrations

, is 4. Calculations show that this value is conserva-j tively applied because the actual ratio is considerably j less than 4. Refer to Subsection 8.3.1.2 4

RG 1.75 " Physical Independence of Electric Systems" F (Rev 2)

The design is consistent with the criteria for physical i independence of electric systems established in Attach-

} ment "C" of AEC letter dated December 14, 1973, and

is in general conformance with Regulatory Guide 1.75, y except as follows

o Battery Room Ventilation. Although the four j Class 1E batteries are housed in separate safety y i class structures, they represent only two redun-dant load groups (see Subsection 8.3.2). Each

load group is served by a separate safety-related ventilation system. There is a cross-tie between j{ the two ventilation systems to allow one system

~ ~ <- ,to serve both load groups in case the other

, system is inoperable. Fire dampers are provided to isolate each battery room. 45'

! For additional information on the four batteries and j two redundant load groups, see Subsection 8.3.2.1.a.

! Refer to Subsection 8.3.1.2.b.5 for a discussion of the onsite ac power system.

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- $ee MSM 1 51

, -RG 1.108 " Periodic Testing of Diesel Generator Units Usad as (Rev 1) Onsite Electric Power Systems at Nuclear Power Plants" i

l~ The diesel generator testing is in conformance with

the-recounsendations of Regulatory Guide 1.108 with one clarification:

The requirements of position C.2.a(5) will be met i_ every 18 months as follows

l-N The L functional capability at full load temper-l ature will be demonstrated at least every 18 months by performing the test outlined in posi-f tion C.2.c(1) and (2) immediately following the f

, full load carrying capability test described

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.-y ~in position C.2.a(3). The full load carrying I

8.1-7

Attachment 4 Page 2 of 5 INSERT I The requirements of position C4, as it relates to cables for the associated circuits, is clarified as follows:

Instrumentation, control and power cables used for the associated circuits will not be covered by the Operational Quality Assurance Program (OQAP). However, programmatic controls will be applied to these items.

The actual implementation of these controls will be defined by the program manuals used to control specific activities at Seabrook Station.

Implementation of these programmatic controls will be verified by Quality Assurance personnel to the extent necessary to insure proper applica-tion. For further details on provisions and considerations for the associated circuits, see FSAR Chapter 8, Section 8.3.1.4.b1d.

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Attachment 4 Page 3 of 5

. SB 1 6 2 Amendment 52 o

FSAR December 1983 hi

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4. Regulatory Guide 1.63 - Electric Penetration Assemblies in Containment Structures for Water-Cooled Nuclear Power Plants l _

46 The electric penetration assemblies are designed to withstand, without loss of mechanical integrity, the maximum fault current 3 vs. time conditions that could occur as a result of single F

F random failures of circuit overload devices. The 600 volt -

system X/R ratio used in specifying the electrical penetrations '

is 4. Calculations show that this value is conservatively applied because the actual ratio is considerably less than 4.

To preclude damage to electric penetrations due to single 3 failures of circuit overload protection devices, each penetra-tion circuit, with the exception of instrumentation and low *gs energy circuits, is provided with dual Class 1E overload - 4

_ protective devices. For more details refer to Subsection  ;

8.3.1.1.c. 15 kV penetrations are protected by seismically

qualified Class 1E fuses. Additional protection is provided r =

by two non-Class 1E breakers in series. These breakers are ,

_ coordinated and derive their control power from different g batteries. For more details refer to Subsection 8.3.1.1.a. El

5. Regulatory Guide 1.75 - Physical Independence of Electric i -

Systems

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The design is consistent with the criteria for physical l j z

independence of electric systems established in Attachment "C" St.

of AEC (NRC) letter dated December 14, 1973. Attachment "C" =

which is incorporated as Appendix 8A, is in general confor-

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mance with Regulatory Guide 1.75.

Physical separation and identification of circuits are <

described in detail in Subsections 8.3.1.3 and 8.3.1.4, 46 47 k respectively. , i

c. Compliance to Branch Technical Position PSB Adequacy of Station Electric Distribution System Voltages Position B1 1.

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. An acceptable alternative to the second level undervoltage protection system described in Position 1 is provided. This alternative system is descibed in Subsection 8.3.1.1.b.4.(b). .

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2. Position B2 The Seabrook Station design meets Position 2 of Branch . ,

R Technical Position PSB-1. The bypass of the load shedding

• feature during sequencing, and its restoration in the event of_a subsequent diesel generator breaker trip, is discussed in -

Fon clari( c'aMon d Pa* ton CM os if veMc5 to st.

cLssocMel circds, r eGer +o FS A R Sec. Hon 1r.l.S.3 b.

[ 1 8.3-37 4 i =1 S

M Attachment 4 Page 4 of 5 SB 1 & 2 Amendment 53 FSAR August 1984 (c) All Non-Class IE protective circuit breakers will be periodically inspected approximately once every five years according to a program developed for the inspection of Non-Class IE equipment. This program will be in accordance with manufacturer's recommendations for main-e tenance and inspections.

_ Since Class IE and Non-Class lE protective devices are i identical, any generic degradation such as setpoint drift, F manufacturing deficiencies, and material defects will be detected and corrected as a result of the rigorous program performed on the Class 1E protective devices to satisfy k the requirements of ANSI N-18.7-1976 and Regulatory Guide

1.63; therefore, credit can be taken for this equipment E to function under DBE conditions.

(d) The probability of an ensuing fire is minimized because

, all cables utilized for these associated circuits are specified, designed, manufactured, and installed to the same criteria as Class IE cables. Factors that have se been taken into consideration include flame retardancy,

[ C C' PUNn3 And non propagating and self extinguishing properties, E fenningfiog reguire Plicing restrictions, appropriate limitations on raceway x ill,pappropriate cable derating, and environmental 7

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qualificationsy w

(e) Degradation of an associated circuit because of a raceway failure during a DBE, has been eliminated because all r electrical raceway systems within the Nuclear Island are r seismically analyzed.

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( f) Other design considerations that contribute to the integrity of these associated circuits are:

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1) Cables associated with one train are never routed in racevays containing Class IE or associated cable of another train or channel.

2) All cables for instrumentation circuits utilize shielded constraction which minimizes any unaccept-4 able interaction between Class lE and associated b ,

circuits.

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3) All circuits entering the reactor containment are provided with prot <sctive devices complying with 7 Regulatory Guide 7.63. For exceptions see p

Subsection 8.3.1.1.C.7(a). l 63 h Based o s the above design features and analysis, we do not M consider these associated circuits to pose any challenges to any Class IE circuits. There fore , the ability for safe plant

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I shutdown under DBE co.4ditions has not been jeopardized.

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$2 n e a bove il?d=3 1h*

g constructionemision3 phase of the ad pln,d consiae (Jtli also diJd'If.A & *duri be used ***udel "".hons 3 the opem phase.

w Attachment 4

'+; Page 5 of 5 SB 1 & 2 Amendment 52 FSAR December 1983

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g.f to normal operation as practical, the full operational sequence L that brings the system into operation, including portions of k -the protection system, is tested, hl Compliance with Regulatory Guides b.

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) 1. Regulatory Guide 1.6 - Independence Between Redundant Standby l

Power Sources and Between their Distribution Systems St.

I The safety-related portion of the station de system for each f unit includes four batteries. The. redundant safety-related

[ load groups are each fed by a separate battery and battery Q charger. There is no provision for automatically connecting j- one. battery-charger combination to any other redundant load

f. . group, nor is there any provision for interconnecting batteries d- either manually or automatically. To further enhance safety E and reliability, two de supply. buses of the same train may be 3i . connected together manually, but circuit breaker interlocks j_ prevent an operator error which would parallel two batteries.

(See Figure 8.3-37).

1. j 2. . Regulatory Guide 1.32 - Criteria for Safety Related Electric Power Systems for Nuclear Power Plants-SL .

The design is consistent with the requirements of this regula-

[ tory guide. For details, refer to Subsections 8.3.2.1.c and v 8. 3. 2.1. e .

[i 3. Regulatory Guide 1.75 - Physical Independence of Electric Systems

{f 51 The design is consistent with the criteria for physical g independence 'of electric systests ' established in Attachment "C" M of AEC letter dated December 14, 1973. Attachment "C" is incorporated as FSAR Appendix 8A and is considered similar to Regulatory Guide 1.75.

4. Regulatory Guide 1.129 - Maintenance, Testing and Replacement of Large Lead Acid Storage Batteries for Nuclear Power Plants

.For compliance to this regulatory guide, refer to Subsection g -8.3.2.1.e.

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c. -Compliance with IEEE-308, Class IE Electric Systems *

( The station de-system conforms to the requirements of IEEE-308.

The power' supplies, distribution system, and' load groups (see Sub-

, section 8.3.2.1) are arranged to provide direct current electric For clarifti 4t'on of pos'tfion- C '1

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<s it veldes Fo associdfed

. cter.uds i reb to = FS AR : Sec.fion g.1.5'.'3 b.

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