ML20087L417

Forwards Supplemental Info in Response to Generic Ltr 83-28 Re Required Actions Based on Generic Implications of Salem ATWS Events

ML20087L417
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	03/23/1984
From:	Murphy W VERMONT YANKEE NUCLEAR POWER CORP.
To:	Vassallo D Office of Nuclear Reactor Regulation
References
FVY-84-25, GL-83-28, NUDOCS 8403270190
Download: ML20087L417 (10)

Text

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,o VERMONT YANKEE-NUCLEAR POWER CORPORATION

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FVY 84-25 RD 5, Box 169. Ferry Road, Brattleboro, VT 05301

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ENGINEERING OFFICE 1671 WORCESTER ROAD FRAMINGHAM, MASSACHUSETTS 01701 TELEPHONE 617-872-8100 March 23, 1984 U.S. Nuclear Regulatory Commission Washington, D.C.

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office nf Mylear pagtor n gulation e

Mr. Domenic B. Vassallo, Chief Operating Reactors Branch No. 2 Division of Licensing

References:

~(a) License No. DPR-28 (Docket Na, 50-271)

(b) Letter USNRC to All Operating Reactors, Generic Letter 83-28, dated July 8,1983

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-(c) Letter, VYNPC to USNRC, FVY 83-117, dated November 7, 1983 (d)

"BWR Scram System Reliability Analysis," W.P. Sullivan, et al, September 30, 1976 (transmitted in letter from GE to NRC, " General Electric Company ATWS Reliability Report," September 30,1976)

(e) Letter. VYNPC to USNRC, Proposed Change No. 79 to Facility Operating License DPR-28, dated March 17, 1980 (f) letter, USNRC to VYNPC, Amendment No. 58 to Facility Operating License DPR-28, dated November 3, 1980

Subject:

Generic. Letter 83-28, Generic Implications of Salem ATWS Events By Reference (b), you requested that we address various concerns resulting from the generic implications of the Salem ATWS Events and pro-

- vide you with our current conformance status, plans and schedules for any needed~ improvements. By Reference (c) we provided you with our initial

. response and stated that additio.nal information would be forthcoming. The purpose of this letter is to provide you with the enclosen supplemental information.

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' U S3: Nucicar R atcry Commission March 23,1984 Page 2 VERMONT YANKIIII NUCLEAR POWER CORPORATION 4

We' trust that this information adequately addresses the subject con-cerns; however, should you have any questions in this matter, please con-tact us.

Very truly yours, VERMONT YANKEE NUCLEAR POWER CORPORATION t*:r -

My Warren P. M phy

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Vice Presi ent and Manager of Operations WPM /dm STATE OF VERMONT ss WINDHAM COUNTY.-)

.Then personally appeared before me, W.P. Murphy, who, being duly sworn, did state that he is Vice President and Manager-of Operations of Vermont Yankee Nuclear Power Corporation, that _he is duly authorized to execute and file the foregoing' document;in the name and on the behalf of Vermont Yankee Nuclear Power Corporation and that the statements th gein are true to the best of his knowledge and belief.

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ENCLOSURE 1 i

VERMONT YANKEE ADDITIONAL RESPONSES TO REQUIRED ACTIONS BASED ON THE GENERIC IMPLICATIONS OF[ SALEM ATWS EVENTS

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Item 1.1 Post-Trip Review - Program Description and Procedure

. _ ' As discussed in Reference (c) of the cover letter, a program to address the intent of-Items 1.1.1 through 1.1.7 was scheduled to be in place by April 1,1984.

In _the interim we have instituted a Standing Order for post-trip review, as

' described-in Reference (c).

At the present t1me we are finalizing new procedures for post-trip review.

We anticipate that these procedures will be approved and implemented by the end of April 1984. A report describing our overall program is being developed and is expected to be finalized in September 1984.

It is our intent to submit this

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report to the NRC_in October 1984. Should our schedule for implementing the post-trip review procedures or submitting the program description change,-we will notify our Project Manager.

Item 1.2 Post-Trip Review - Data Information Capability

. A_ report describing our existing data and information capability was sub-

-mitted via Reference (c). As discussed in that report, modification to our capability may result from our programs to address NUREG-0737, Supplement I; however,-no other changes are being considered at this time.

Item 2.1 Equipment Classification and Vendor Interface A.

Reactor Trip System Components Identified as, Safety-Related As described in Section 3.1.2.5 of NUREG-1000, the GE Boiling Water Reactor (BWR) trip system design differs from the Pressurized Water Reactor

.(PWR)' design.'.The GE. reactor trip system consists of redundant plant pro-cess instrumentation that feed one out of two twice logic that initiates a reactor trip by de-energizing solenoid operated scram pilot valves which vent' air from the scram valve diaphragms and insert the-control rods.

These components are contained within several functional systems at our plant; rather than one system called a Reactor Trip System (RTS). The systems which provide the reactor trip function are as follows:

SYSTEM NAME DESCRIPTION (a) Control Rod Drive Scram valves, scram discharge volume water level sensors, backup scram valves d

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Enclosurg I Page 2 (b).ReactorProtection logic, power supplies, turbine sensors, drywell pressure sensors (c)' Neutron Monitoring Neutron flux sensors, trips and hypasses (IRM, APRM)

(d) Nuclear Boiler Reactor high pressure sensors, reactor pressure (600 psi bypass) sensors, main steam line isolation valve sensors (e) Process Radiation Main steam line radiation sensors Monitoring It' is not aur intent to establish a new system called the Reactor Trip System. We believe that creating a "new" system.made up of sub-sets of components from existing systems will result in confusion and necessitate

.the development / revision of plant procedures, manuals and drawings.

Our review to ensure that the components whose function is required to trip' the reactor are identified as safety-related on documents, procedures and information systems used in the plant to control safety-related activi-ties.,Lwill be included in our review of all safety-related systems, as discussed in our response to Item 2.2.1.

B.

Reactor Trip System _ Vendor Interface Program In response to this concern, Vermont Yankee joined with 55 other uti-lities and formed an INP0 Nuclear Utility Task Action Committee (NUTAC).

This committee has developed and approved an industry-wide Vendor Equipment

,Technica1'Information Program (VETIP). This program is described in detail in a final report scheduled to be submitted to the NRC in the near future.

The program promotes interaction among the major organizacions involved in the generation of commercial nuclear power. Under the program, individual utilities exchange safety-related systems and component information with vendors, the-NRC, INPO and other utilities. This exchange takes place via written notifications (i.e., Licensee Event Reports, NRC 18E Bulletins and Information Notices, industry newsletters, etc.), as well as industry meetings and day-to-day comunication. The goal of these mechanisms is to share equipment technical information so as to improve the safety and reliability of nuclear power plants.

The primary purpose of the VETIP program is to ensure that current information and data will be made available to those personnel responsible for developing and maintaining plant instructions and procedures.

It should be noted that these information systems and programs currently exist and are capable of identifying precursors to the industry that could lead

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' Enclosure I L

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to a Salem-type. event. The VETIP-program formalizes the exchange of infor-

- mation;through an industry controlled and primarily hardware oriented program that'.does. not rely on vendor action (other than the NSSS Supplier)

Lto provide information directly to.the. utilities.

Instead, the VETIP

program'provides informationLdeveloped by industry experience. through y

Significant Event-Reports -(SER's), Significant Operating Experience ReportsE(SOER's) ~and 'other similar reports to the appropriate equipment y

vendor.for comment before it is circulated to the utilities concerned.

. Vermont.. Yankee has an existing vendor equipment information program

.with General Electric-(GE), our NSSS vendor. This program consists of two major categories: :(1) information regarding' safety-related systems and components; and (2). technical information intended to enhance safety and h'

non-safety-related. equipment reliability and improve plant performance.

The' programs include, but are not: limited to:

j(a)f10CFR21 Reporting The General Electric' Comp'any has established a ' reporting system to handle ' safety concerns that complies with the requirements of 10 CFR Part 21'.-

(b)f Urgent Coramunications In addition to the 10 CFR 21-reports, a procedure for handling urgent communications to BWR owner / operators has-been established for use in M

providing prompt notification of. safety concerns. These com-munications'usually take the-form of a-short: letter which provides a 4.

' brief' explanation of the problem and advice or precautionary measures to be observed. to avoid. potential operational hazards. ~ Due to their urgent ' nature,' these communications :are processed to operating plants

-by the most effective method. (i.e'., telex, telecopy, cable, special

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mail handling, 'etc.) ~and are normally followed up by a telephone call-b<

~to~aipre-designated utility individual to assure receipt of the infor '

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mation.

, (c). Service Information Letters (SILs)

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- These -informaticn : letters usually provide recommendations for-equip-ment. modifications, plant design improvements or changes to procedures

~to improve p1 ant performance.

(d);; Service Advice' Letters-(SAL's) b~

These letters are -issued by the GE Product Departments -(other than the

San -Jose, CA based Nuclear Engineering Product Department) and are used.to provide. notification of product problems and for service

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Y information on a' broad range of consumer and industrial products.

tThose SAL's that are recognized.by the issuing Product Department as f

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1 applying to. devices / equipment used in nuclear-power plants are spe-A cially 1dentified and are flagged for distribution to all nuclear -

Lpower plants.'

(e) Turbine Information Letters (TIL's)

The letters are issued by GE's Large ' Steam Turbine Generator Department to Lprovide descriptions of product problems / improvement and to recom-imend modifications that will mitigace problems and/or improve product

- performance.

= 0ur involvement in the NUTAC VETIP program will provide a mechanism to Jassure 'our recaipt of.all applicable information from GE.

As discussed in our response to Item 2.2.2 in Reference (b), we have a program in place to

assure that the' above information, _as well as' any other information

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received which is' pertinent to plant operation,1s. effectively reviewed,

/ assessed, distributed and acted'upon by appropriate personnel.

This program 1sidefined in a plant procedure which was established in l response;to:NUREG-0737-Item I.C.5,. Operating' Assessment, and incorporates

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cthe -review of information for both safety-related and non-safety-related equipment and components.

D We believe'our existing. program, as'well as plans to incorporate the VETIP program (a3' discussed in the our response to Item'2.2.2), adequately Laddress.-your concerns with-respect to ' receipt and d.isposition of vendor

. equipment information.

, Item 2i2 Equipment Classification and Vendor' Interface ~ (Programs for All Safety-Related Equipment)

Item 2.2.11 Equipment Classification As ' discussed in Reference (c) of the cover letter, a program exists ifor determining-safety classification in accordance with ANS-22,

' Draft No.f 4, Revision:1,' May 1973, as described in the NRC-approved Yankee 4

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. Atomic Electric. Company Operational' Ouality Assurance Manual (Y00AP-1-A).

The. general classification of structures, components, and systems is deli-

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t (neated in Appendix C of YOOAP-1-A, with further breakdown prcvided on

system flow diagrams and electrical' one-line diagrams.

R iBy_ Reference-(c), we informed you that we were reviewing the adequacy

-off our existing program and participating with the BWROG in the resolution 1

? of;certain aspects of'this item. 'We have since determined that, although

' - Jour-existing program meets all applicable criteria. programmatic enhance-

.mentsishould be implemented.

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s j This.enha'ncement will' requireithat.we re-review all safety-related Lsystems.to ensure that the components (including associated instruments, controls and electrical equipment) necessary for a particular safety.

related system to perform its safety function are classified as safety-related on-plant documents,; procedures and information systems.

It is our present:intentito structure our re-review such that'those systems which

comprise the: reactor' trip function,;as identified in our response to Item 2.1, will'.be addressed first.

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We 'are presently developing an' outline for this. enhanced program and expect to have the program: defined, in-terms of scope and manpower needs,.

.by -September 1984. ' At that time we will inform you 'of our plans and sche-

dule foricompleting' this effort.

It should be noted that this program will

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undoubtedly require an extensive allocation of engineering resources and Lits development / completion will have to be integrated with ongoing en-

gineeringactivities.

'With respect to the equipment classification program in use at Vermont Yenkee for structures, systems and components Important to Safety, we are participating in the Utility Safety Classification Group and are seeking a

' generic resolution to the Staff's concerncin.this regard through-the efforts of.the Group.-..Our current position. remains as stated in our origi-nal: response to Item 2.2.1.6.

? Item 2.2.2 Vendor Eduipment-Interface. Program-Asidih ussed.in our response to-Item 2.1.(B), we participated in the INPO

-NUTAC effort to develop the Vendor Equipment Technical Information Program

(VETIP).

A' final report' describing this program is scheduled to be submitted to the NRC in the'near-future.c It 'should' be noted that since the program was deve-

-loped on 'a' generic basis-to enhance existing plant vendor information programs, certain provisions _ of the VETIP. may not he applicable to some utilities or may.

irequireimodification to fit.into.the plant specific equipment information

-.progra'm.-(Thus','we expectfit will take 90 days from our formal receipt of the

- final Lreport!from.INPO to determine the scope of our. final program and to Lestablish a schedule to its implementation. ~At that time'we will inform you of

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.our.expecte'd irapiementation date. for this program, t.i !

L item 3.1 Post-Maintenance Testirig '(Reactor-Trip System Components)'

I Items'3.1.1 through 3.1.3 were. addressed in Reference (c)..

JItem 3.2IPost-Maintenance Testing (A1110ther Safety-Related Components)

(Items 3.2.1 through 3.2.3 were addressed in our response to Items 3.1.1

.through_3.1.3, as discussed.in Reference (c).-

Item =4.1 Reactor Trip System Reliability (Vendor-Related Modifications)

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This item is not applicable to Coiling Water Reactors.

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Item 4.2 Reactor Trip System Reliability (Preventive Maintenance and Surveillance Program for Reactor Trip Breakers)

This item:is' not applicable to Boiling Water Reactors.

Item '4.3 Reactor Trip System Reliability (Automatic Actuation of Shunt Trip Attachment for Westinghouse and B&W Plants

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This item is not applicable to Boiling Water Reactors.

Item 4.4 Reactor-JTri System Reliability (Improvements in, Maintenance and

Testing Procedures for B8W Plants)

This item-is not applicable to Boiling Water Reactors.

Item 4.5.1 and 4.5.T Reactor Trip System Reliability (System Functional Testing)_

As disc'ussed in Reference '(t) of the cover letter, Section 3.1 of our faci-

1. l i ty : Technical Specifications provide' functional test surveillance requirements

' for the Reactor Protection System (RPS). This includes a requirement to perform Jon-line' functional testing of the RPS, including indeper. dent testing of. the scram-pilot valves.

Item 4.5.1 of the Generic Letter also recommends on-line

-functional-testing of the backup scram valves at GE plants.

The backup scram valves are, implied to be " diverse trip features" comparable to the breaker shunt trip 1 features on other plants and, as such, 'should be functionally tested on-

line.

However, we believe the inherent difference between the GE RPS design and the Reactor Trip System design of-other plants, makes the recommendation for on-line. functional-testing of. the backup scram valves unwarranted.

Specifically, Vermont. Yankee has 89 control rods. Each control rod is activated by. a pair of independent scram pilot valves.. Reference (d) of the cover letter is an analysis performed by GE which concluded that reactor shut-

'down can be achieved if at least 50% of the control rods in checkerboard pattern or 69% of the control rods in a random pattern are inserted into the core.

Clearly, only a fraction of these 89 control rods must successfully function to 1 shutdown the: reactor. The probability of independent failure of enough rods to 4

prevent shutdown is negligible.

. Two' re'dundant backup scram valves are provided in GE plants to assure that the control. rods.do actuate should any of the pilot scram valves fail to func-

-tion. No explicit credit is taken for these valves in plant safety analysis or system reliability analyses, nor are they required by applicable regulatory Trequi rements.- Functional. testing of these valves during plant operation would require a plant scram,- a significant challenge to plant safety systems, and therefore a degradation in plant safety. The backup scram valves are non-safety-related additions that can only enhance the reliability of the safety-related Reactor Protection System.

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e Enclosure I Page 7 In contrast, other reactor designs include only two redundant reactor trip

. breakers, one of which is required to successfully function to scram the reactor

'Each of these. breakers has an undervoltage actuation device and diverse

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functionally redundant shunt actuation devic 2.

-The successful function of this e

system requires-the' operation of 1 of these 4 actuation devices.

In the event of an initiating event-(loss of AC power) followed by a single active failure (breaker _ failure), the shunt devices are rendered useless, therefore successful system operation depends'on breaker operation by the remaining action device.

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!Obviously, improper function of the; diverse safety-related trip devices on other plants would degrade the reliabil.ity of the safety-related Reactor Trip

_ System.

e In summary, the functions performed by and the safety-related reliability

. dependence on shunt actuation devices in one design are considerably different from the reliability enhancement afforded by backup scram valves in the GE

. design.

Similarly,-testing requirements on shunt actuation devices should not necessarily be requirements-for backup scram valves. We therefore conclude that the need for on-line functional testing of the backup scram valves is not necessary.

'It.should be noted that'we have also installed an Alternate Rod Insertion -

.(ARI) system and a Recirculation Pump Trip (RPT) system in response to previous

-ATWS concerns. The ARI system provides a means to reduce the probability of a

-failure of control rods to insert on demand, while the-RPT system is provided to

. mitigate ~ the consequences of an ~ATWS by reducing core power generation. by rapidly' reducing core flow. The. parameters which initiate the ARI and.RPT systems ~are reactor low-low water level, after a time delay of approximately ten seconds, or.high reactor pressure. The use of a ten-second delay on low-low water level trip is desirable to avoid making the consequences of a post'ulated loss-of-coolant accident more severe.

.These' systems are' described in detail in Section 7.18 of our Final Safety

_. Analysis'. Report 1(FSAR).. The installation of the RPT system is also reflected in

-our. Proposed Change:No. 79 to facility Technical Specifications (Reference (e)),

-which was subsequently approved-by the NRC~in Reference (f).

Item 4.5.3-Reactor Trip System Reliability (On-Line Functional Testing Intervals)

- As d.iscussed above, our facility Technical Specifications provide func-tionalitest/ surveillance requirements for the Reactor Protection System. These,

as well.as all other Technical: Specification requirements, are continuously reviewed such that~ test intervals are consistent with achieving high component andl system're11 ability.

If existing intervals are found to be inconsistent with

. achieving high reliability, we would propose a change to our Technical

-Specif'ications to modify them accordingly.

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g Enclosure I Page 8 In addition,.we.are following the efforts of the BWR Owners Group Technical

. Specification Improvement Committee; which is addressing this particular item as

'part of theirLoverall effort to review and recommend improvements for various equipment test / surveillance intervals. At the present time, the BWROG committee thas 'not' established a ofinn schedule for completing this effort; however, once a final' report is made, it:is our intent to review the applicability of their recommendation for.this particular item as it relates to our facility _and sub-mit.a -fonnal request to amend our. Technical Specifications, as necessary.

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