ML20087E471

Forwards Minutes from 831205-08 Meeting W/Instrumentation & Control Sys Branch Re Discussion of 830919 Request for Addl Info.Items Listed as Closed Imply That Info Requested by NRC Has Been Provided

ML20087E471
Person / Time
Site:	Beaver Valley
Issue date:	03/09/1984
From:	Woolever E DUQUESNE LIGHT CO.
To:	Knighton G Office of Nuclear Reactor Regulation
References
NUDOCS 8403160190
Download: ML20087E471 (21)

Text

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/gK 2NRC-4-026 (412) 787 - 5141 Telecopy 8-6 Nuclear Construction Division March 9, 1984 Robinson Plaza, Building 2. Suite 210 Pittsburgh, PA 15205 United States Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Mr. George W. Knighton, Chie f Licens ing Branch 3 Of fice of Nuclear Reactor Regulation

SUBJECT:

Beaver Valley Power Station - Unit 2 Docket No. 50-412 ICSB Meeting Minutes Gentlemen:

Your le t t e r of September 19 , 19 83, fo rwarded " Safety Review Reque s t a for Addit ional In fo rma t io n" from the Ins trument ation and Control Systems Branch (ICSB). Duquesne Light Company (DLC) met with the ICSB on Decembe r 5-8, 1983, to discuss their que s t ions cont ained in your le t t e r.

The enclosed minutes of that meet ing were prepared by DIE and have be en reviewed by the ICSB reviewer. Those items lis t ed as " closed" are under-s tood to be closed only in the se ns e that the information requested by the ICSB reviewer has been provided.

DU t SNE LIGHT COMPANY 6LLp By glA . * ~

EJ J . Woolever Vice President KAT/wjs At t achment cc: Mr. H. R. Denton, Director NRR (w/a)

Mr. D. Eisenhut , Director Division of Licensing (w/a)

Mr. G. Walton, NRC Resident Inspector (w/a)

Ms. L. Lazo, Project Manager (w/a) 8403160190 840309 PDR ADOCK 05000412 A PDR

1 I

I MINUTES OF DECEMBER 5-8, 1983 MEETING BETWEEN THE NRC INSTRUMENTATION AND CONTROL SYSTEMS BRANCH AND DUQUESNE LIGHT COMPANY l

1. Identify any plant safety-related sys tem or port ion thereof, fo r wh ich (7.1) the design is incomplete at this time.

Response

The fo llowing is a list of plant safe ty-r ela t ed sy s t ems o r po rt ions thereof, for which the design is presently incomplete:

1. Plant Safety Monitoring System (includes ICC ins trument ation and port ions of the Reg. Guide 1.97 ins trument ation)

2. Pressurizer Relief and Safety Valve Indication System

3. Main Steam Radiation Monitoring System

4. Control Room Isolation and Pressurization System

5. Charging Pump Miniflow Isolation Valves Status:

Closed

2. As called fo r in Section 7.1 of the Standard Review Plan, provide infor-(7.1) mation as to how your design conforms with the following TMI Act ion Plan Items as described in NUREG-0737:

(a) 11.D.3 - Relief and safety valve position indication (b) II.F.1 - Accident monitoring instrument ation (Subpart s 4, 5, and 6)

Response

a) FSAR Sect ions 1.10 and 7.5 record the commi tme nt to incorpo rat e s afe ty-rela t ed pres surizer powe r-ope ra t ed e-lief valve pos it ion ind ica t ion.

b) Subpart s 4 and 5: FS AR Table 7.5-1, and Sect ion 1.10 were refe r-enced indicating compliance with NUREG-0737.

Subpart 6: The H 2 Analyzers are used to monitor hydrogen concen-t ration ins ide cont ainme nt pos t ac cide nt . The anal yze rs and .

their piping system, controls, indications, accuracy and qualifi-cation meet the requirements of NUREG 0737 II.F.1, Attachment 6.

The analyzer system is automat ically initiated within thirty minutes af ter an initiating event from accessible control panels.

Status:

a) Confirmatory; based upon- filling in the "l a t e r" info rmat ion in Tab le 7.5-1.

b) Closed  !

l

3. Provide a brief cv e rv i ew of the plant electrical di s trib ut io n sys ts,

( 7.1 ) with emphasis on vital buses and separation divisions , as background fo r addressing various Chapter 7 concerns.

Response

he basic electrical plant di s trib ut io n sys tem was di sc u s sed ref e r-encing FSAR Section 8.3.1 and Figure 8.3-3 which depicted the "Vi tal Bus Sys tem One Line." The discussion demons trated the arrangement of the two-tra in/ four-ch annel red unda nt Class IE uni nt errupt ib le powe r supply system including its operability and associated load groups.

Status:

Closed

4. Describe design cri teri a and tes ts performed on the isolation devices in

( 7.1 ) the Balance of Plant Sys tems . Address results of analysis or tes ts per-fo rmed to dennns tra te pr ope r isola tion be twe en sepa ra tion gr ou ps and between safety and nonsafety systems.

Response

A description of the isolation devices used for digi tal signals was pr es e nt ed . The qualification aspects of the devices was di sc us sed .

Tes t reports are available to verify the qualification anl design of the devices.

The discussion also described the voltage regulating isolation trans-fo rme rs th a t are used to provide for Class IE to Class IE and Class 1E to non-Class IE i sola t ion . Fac to ry tes ts were de sc ribed th at demons trated the transformer's ability to provide for isolation.

Status:

Closed 5.

(7.1) Describe fe a tures of the Beave r Valley 2 environnent al control sys tem wh ich ins ure th at ins trumantation sensing and sampling lines for sys tems impo rt ant to safety are prot ect ed from fr eezing during extremely cold we a the r .

Discuss the use of environmental moni toring and alarm sys terns to pr event loss of, or damage to , sys tems impo rt ant to safety upon failure of the environmental control and monitoring system circuits.

Response

The Class IE electrical heat traci ng sys tem was discussed including its ability to monitor each line and provide control via the associ-a ted RTD . The discussion included mention of the local and remote annunci a t ion including the ability to re ad actual temperature mea-surements for each circuit. FSAR Section 8.3.1.1.3 was referenced.

Status:

i Closed

6. Provide a list of any non-Class IE control signals that provide input to (7.1) Class IE control circuits.

Response

The requested list was provided Status:

Closed

7. Identify Were mic roproces so rs , mult iplexe rs , or comput er sys tem s are (7.1) u sed in or interface wi th saf e ty-rela ted sys tems . Also identify any "first-of-a-kind" ins truments used for safety -related systems.

Response

For radiation monitoring, BVPS-2 is utilizing a microprocessor for e ach safe ty-rel a ted radiation mo ni to ri ng vari ab le . The mic ro-proce s so r is clas sified Class IE and quali fied in ac cordance wi th IEEE 323 for the environnent in s ich the microproces sor is loca ted .

There are fourteen microprocessors for Category I radiation monitors .

The vendor for the BVPS-2 radiation monitoring sys tem is GA Technol-ogies, 40 h as prov ided these monitors to a variety of other nuclear power plants.

The multiplexers used in the SSPS are similar to those used on previous plants.

In addition, BVPS-2 also has a firs t-of-a-kind processor based Plant Safety Monitoring System. Section 7.7 of the FSAR was referenced.

Status:

Closed

8. We reque s t th a t the se tpoint methodology for each Reactor, Protection

( 7.1 ) Sys tem (RPS) and Engineered S afeguard s Features (ESF) trip' setpoint values be provided for both NSSS and BOP scope of supply at' the time - the Technical Specifications are submitted for review.

Response

A description of the setpoint methodology will be provided den the Technical. Specifications are submitted.

Status:

= Confirmatory: based upon the' submittal ~of , the methodology.

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9. Ident ify any Balance of Plant sc ope safety-related equipment (othe* than (7.1) those listed in Section 7.1.2.4 of the FSAR) that cannot be tes ted during reactor operation. ' Include auxiliary relays or other- component s in the safety-related systems.

Response

FSAR Sect ion 7.1.2.4 lista the non-t es table equipment . No ot her non-testable equipment has been identified.

Status:

Closed

10. In Section 7.1.2.6 of the FSAR compliance with R.G. 1.53 addres ses only (7.1) prot ect ion sys tems . Provide the equivalent information for other systems important to plant safety.

Response

The FSAR confirms compli ance with single failure cri teri a fo r all safety-related equipment. ,

Source: .

Closed

11. Discuss the following:

(7.1)

(a) Response time t es t ing of" BOP - and NSSS - prot ect ion systems using the design criteria described in position C.5 of R.G.1.118 and Sect ion 6.3.4 of IEEE 33fi n (b) Identify any temporary j umper wires or tes t ins trume nt at ion diich will be used. Provide further

• discussion to describe how the t es t procedures for the protection systems conform to' R.G.1.118 position .

C.6.

(c) Typical res pons e time t es t methods for pr es sure and t empe rature sensors.

(d) Co:npliance with St andard Technical Specifications for Wes t inghouse

. Pres surized - Water t Reactors -(NUREG-0452, Rev. 4) as rela t ed to - the third paragraph _ of

• the discussionunder R. .G. 1.118 on page - 52. of FSAR Table 1.8-1. c ,, ^\ .

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Response

a) This - was '. discus se'd. as ' indicated -in the ' FS AR pos ition on R'.G.

1 .118 and the ; Pre-operational- ' Test Pr ogram - Tes t ' Abstracts

14. 2.12. 2. 2 and 14.1 ~.12. 2.4. , s

-b) No - _ tempo ra ry j umpe r- wires wil l' . be t ut ilized during operational

. t es t ing of l the prot ect ion sys tem. .

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E1 c) The response time tes t method planned to be utilized for tempera-ture se nso rs is the Westinghouse analys is methods currently in use on Beaver Valley Unit 1.

d) Beaver Valley is in compliance with the S t andard Technical Speci fica t ion.

Status:

Closed; t empe rature se nso r res pons e t ime test me tho ds have be en ide nt ified .

12. Using de t ailed plant design drawings, discuss the reacto r trip breake r (7.2) and undervolt age relay tes ting procedures and the capability of indepen-dent verification of the ope rability of react o r trip breake r shunt and undervoltage coils.

Response

This information is not yet available pending the development of the response to Generic Letter 83-28 (due April 1,1984).

Status:

Confirmatory; pending the transmission of the response to G.L. 83-28.

13. Using de t ailed plant design drawings, discuss the reactor coolant loop (7.2) isolation design and valve interlocks.

(7.6)

Response

A discussion of the various interlocks on the reactor coolant pu.ap and loop isolation valves which are utilized in bringing an inact ive loop into se rvice was prov ided . There - was some discussion of the increased probability of prot ect ion system ch alle nges due to 1/ 2 logic (ve rs us 2/3 logic) when the ins trument ation in the iso la t ed

. loop is. placed in-the trip mode . A summary of the actions taken in the reactor - coolant and protection systems when going to N-1 will be-

. s ubmi t t ed .

Status:'

Confirmatory; pending the submittal of the N-1.information.

14. Table ' 7.2-4 provides ' reactor trip correla t ion for reacto r trip s ign al ,

(7.2) accident analys is , aml ' technical spe ci ficat ions . Ple ase prov'ide ' a (7.3) similar table for safety interlocks and bypasses'.

Response

The safety interlocks and bypasses ~ date is dependent upon .the technical -

specifications which will be generated in mid ~ 1984. A t ab le similar to Table 7.2-4 will be provided to the ICSB in response to this question.

Status:

Closed

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15. Dcacribe the steam generator level instrumentation. Identify the ins tru-( 7.2) ment ch annel used for protection functions and the control fmetions .

( 7. 3) Address the control and protection interaction confot1 nance to Section 4.7 o f IEEE S td . 279-1971.

Response

i ne steam ge ne*a to r level ins trument ation was described. The NRC l expressed concern that there was not adequate protection agains t the controlling ch annel failing low and res ult ing in overfill of the steam generator.

Status:

se Open; DLC will study the matter and respond at a later date

16. Using de tailed schema t ic s , describe the design of pressurizer PORV con-( 7.2) trol and the ' block valves control, and verify that no sirgic failure (7.6) will praclude the automatic actuation logic for all modes of operation.

/

Response

'lh e control of th.e pres suri zer PORV's was di scus sed . The NRC expres sed concern ' that the protect ion portion of the interlock did not appear to be single-failure proof and therefore migh t not be able to isolate flow in the event of the degredarion of the control grade pressure transmitters.

S t:.cus':

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17. We informe* lon in Section -7.2.1.1.2 for " Reactor Trip on a Turbine Trip"

( 7.2) is insu f fis ent. Pleaae provide further design bases discussion on this subject, per/BTP ICSB 26. requirements. As a minimum you should:

- i (1) Using detailed drawings, describe the routing and sepration for this trip circuitry from the sensor in the turbine building to the final actuation, in the reactor . trip system (RTS).

(2) Discuss how , the routing wi thin the nonseismic Catego ry 1 turbine 1 building .is such that " the ef fects of credible faults or failures in i this area on f these circuits will not ch allenge the reac tor trip system aYid thus degrade ~ the RTS performance. 'Ihis should include '.a discussion of, isolation devices.

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(3,t Describe the power supply arrangement

for the reactor trip on turbine trip circult'r'y. -

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(4) Discuss the( tes ting / p1 nned for the reactor _ trip on turbine trip circuitry. [ ,,,g-f-  ! l

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(5) 1. > cuss kualifi5Irio'n' of the sensors.

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Identify other sensors or circuits used to provide input signal s to the other prot ec t ion sys tems s ich are loca ted or routed thr ough nons e is-mically qualified structur es . This should include se nso rs or circui ts providing input for reactor trip, emergency safeguards equipment such ss the auxili ary feedwater sys tem, and safety grade interlocks. Ve ri f ica-tion should be provided th a t the se nso rs and circuits meet IEEE-279 nd are seismically and environment ally quali f ied . Tes t ing or anal yses performed to insure th at failures of nonseismic structures , mountings, e tc . will not cause failures sich could interfere with the operation of any other portion of the protection system should be discussed.

Response _:

A detailed discussion of the re ac to r trip on a turbi ne trip was pr es e.at ed using de t ailed dr owi ng s . Included in the discussion was the potential degradation of the reactor protection sys tem, the power supply arrangement, and the train sys tems with 2 out of 3 logic. It was shown that the technical specifications provided agreed with the standard technical specifications.

Status:

Closed

18. Identify 4ere ins trume nt senso rs or transmi t ters supplying infomation

( 7.2 ) to more than one prot ec t ion ch annel are loca ted in a common ins trume nt (7.3) line or connected to a cmmon ins trument tap. The intent of this its is to verify that a single failure in a common ins trument line or tap ( such as a break or blockage) cannot de fe at req uired pr ot ect ion sy s tem redundancy.

Response

The ins trument sensors supplying information to more than one protec-tion di annel loca ted in a comman ins trument line or connected to a common ins trument tap consis ts of loop flow ard pres surizer pres sure and level instrumentation per the standard Westinghouse design.

Source:

Closed

19. Discurs the method of redundantly tripping the turbine following receipt (7.2) of reactor protection signals requiring turbine trip.

Response

De tailed electrical sch ema tic s were used to desc ribe the re ac tor protection system initiated turbine trip.

Status:

Closed

20. As discussed in Section 7.2.2.3.1 of the FS AR, an isola ted ouput sign al

( 7.2) from protection system ch annels is provided for automa tic rod control.

Discuss how this signal is derived. Discuss d at steps, if any, are taken to prevent unneces sary control action during tes ting of protection system channels with a test source.

Response

NIS powe r range detector tes ting imposes the tes t sign al on the exis ting flux signal . As shown on Fig. 7.2 sheet 9, any power range signal going high because of tes t will irh ibi t manual or au toma tic rod withdrawal. The NIS w ign al uacd for this function is derived from the NIS through an isolation device as shown on the NIS block diagram.

Status:

Closed

21. Discuss surveillance of the RTD bypass loop flow ind ica tio ns . Confirm

( 7.2 ) th a t technical speci fica tions will include surveillance requirement s for these indications.

Response

The RTD bypass loop flow indication will be tes ted every refueling outage.

Status:

Closed

22. Recent review of Waterford revealed heaters were used to control tempera-( 7.2 ) ture and humidity within insulated cabinets housing electrical tr ans mi t-ters th a t provide input to the RPS. These heaters were unqualified and concern was raised that heater failure cc,1d cause transmitter degrada-tion. Please ad dr es s any similar ins talla tions at BVPS-2. If heaters are used, describe design criteria.

Response

He aters are not used to control tanpa ra ture aM humidi ty wi thin insulated cabinets at Beaver Valley 2.

Status:

Closed

23. Using de tailed plant design drawings, discuss the control roon isola tion (7.3) and pressurization systems.

Response

Al though the design of these sys tems is inccmple t e , the concep t for the design was discussed.

E t i g 1 Status:

Open: pending the finalization of the design.

24. Using de t ailed plant design drawings, discuss the containnent as toma tic (7.3) isolation sy s tem. No radi ation signal was shown on the logic di agr am.

Please addres s the diversity requirement stated in Standard Review Plan Section 6.2.4. Also discuss diich valves are pr es ele ct ed for manual operation stated in Item 14 of FSAR Section 6.2.4.1.

Response

FSAR Table 6.2-60 was used as a focal point to discuss containnent isola tion fe a tur es . We NRC requested that Duquesne Ligh t ident ify those pieces of BOP equipment diich pe rfo rm sys tem protect ion fun c-tions but do not perform ESF functions.

Status:

Confirmatory; pending the submittal of the requested data.

25. Using de tailed sys tem sch ema tics , describe the seque nce fo r au toma tic

( 7.3) ini tiat ion , ope ra tion , reset, and control of the auxili ary fe edwa ter (7.4) system. We following should be included in the discussion:

(a) the ef fects of all switch positions on system operation, (b) the ef fects of single power supply failures including the ef feet of a power supply failure on auxiliary feedwater control af ter automatic initiation circuits have been reset in a post accident sequence.

(c) any bypas ses wi thin the sys tem including the me ans by diich it is insured that the bypasses are removed.

(d) initiation and annunciation of any interlocks or automatic isola tions that could degrade system capability.

(e) the safe ty clas sifica tion and design cri teri a fo r any air sys tems required by the auxiliary feedwater system. Bis should include the design bases for the capaci ty of air reservoira - required for sys tem operation.

( f) design fe atur es provided to tenninate auxiliary fe edwa ter flow to a steam generator af fected by either a steam line or feed line break.

(g) sys tem fe atures as sociated wi th shutdown from - outside the control room.

Response

Schematics of the Auxiliary Feedwater Sys tem were presented. Startup of the auxiliary feedwa ter motor. driven pumps and turbine driven

i 1

pumps along with operation of the auxiliary feedwater Control Valves was discribed.

Status:

Closed

26. Using de tailed plant design dr awing s , illus tra te that the canponents

( 7.3) in the auxiliary fe edwa ter turbi ne-driven pump fluid pa th s are to t al ly (7.4) independent from AC power sources . Discuss the capability to control or terminate auxiliary feedwater flow under a loss of AC power event.

Response

Detailed drawings were used to descirbe the following design. The turbine driven pump will operate in the ab se nce of ac powe r . Stean will be delivered to the turbi ne by fail open sole noid valves in branch piping from the main s tean headers . Auxiliary feedwater will be delivered to the stean generators through the auxiliary feedwa ter control valves s ich are normally open. Stean will be removed from the steam generators via the main steam safety valves.

All safety rela ted ins trument a tion is powe red fr om the de powe r system and would be available for at least two hours.

Status:

Closed

27. Discuss the water source s of the auxili ary fe ed wa ter sys ten and the (7.3) capability to transfer one source to the other.

( 7.4)

Response

Detailed drawings were used to de sc ribe the auxili ary fe edwa ter sources from the primary plant demineralized water storage tank and the demineralized water s to rahe tank. In ad di tion, the eme rgency supply from the service water system was described.

Status:

Closed

28. For main s te en and feedwa ter line valve ac tua tion , describe control

( 7.3) circuits for isola tion valves and include au toma t ic , manual and tes t f e atures . Indicate Ae ther any valve can be manually ope ra ted and indicate specific interfaces with the safety system electrical circuits.

Response

Detailed drawings were used to discuss main stean and feedwater valve actua tion. During the review some incons is tencies were found in the drawings describing fe edwa ter i sola tion. Addi tionally it appe ared that the ~ low T avg. reac tor trip fe edwa ter isole tion -was not redundant.

Status: '

Ope n- pending resolution of the drawing incons is tenci es and pe nding DLC providing an adequate description of the low T avg. reactor trip feedwater isolation.

29. Using de t ailed sch ema tic s , describe the ope ra t ion of the cont airunent (7.3) de pres suriza tion sys tem ini tiating circuits, bypas ses , int erlocks , and functional testing.

Response

The de signs of the que nch and recircula tion spr ay sys tens were discussed in de tail using P&ID's to describe the fluid sys tem design and the recirculation spray sys tem interface with safety inj ection (ECCS function). Electrical schematics were reviewed to describe the actuation circui ts for pumps starts and the automatic valve ac t ions for system operation.

Status:

Closed

30. Using logic and schematic diagrams, describe the safety injection systen (7.3) initiating circuits, bypasses, interlocks, and functional testing.

Response

The design of the safety injection systems were discussed in detail .

Status:

Closed

31. Using logic and sch ema tic di agrams , desc ribe ' the AC eme rgency powe r

( 7. 3 ) sys tem (diesel generators and seque nce r), ini tiating circui ts, bypas ses ,

interlocks, and functional testing.

Response

The basic electrical' power arrangement of the red undant Class lE diesel generator units was described as per FS AR Section 8.3. Sch 2-matics were used to show the Emergency Diesel Generator Sequencer and the load shedding schemes.

Status:

Closed

32. As discussed in Section 5.4.15.2 of the FSAR, the < reactor ves sel . heal (7.3) tes t sys tem consists of two parallel . flow paths with redundant isola tion valves in each flow pa th. Discuss ope ra tion of this sys ten from . the

control room. Since the redundant valves are powered from the same vital 1

power supply, discuss dat measures (separation, grounded shield le ais ,

etc) are used' to satisfy item A(8) of II.B.1 of NUREC-0737.

Response

l Sch ema tics for the Reactor Head Vent Sys ters Valves 2 RCS*SOV200A&B ,

l 201A&B. and 2RCS&HCV250A&B were present ed . We power source s ich l supplied each valve was identified indicating that all valves in each flow path were operated from the same - power train. We rationale for this design was discussed.

Status:

Closed

33. Using detailed drawings, describe the ventilation systens used to support (7.3) engineered safety fe atures areas including areas containing sy s teps required for safe shutdown. Discuss the design bases for these sys ters including redundency, testability, etc.

Response

A discussion of . selected ventilation ayu tems used to support engi-neering safety fe atures areas,- including areas containing sys tems required for safe shutdown, was provided at the ICSB meeting. We discussion focused on the design bases fo r the sy s tens including redundancy and tes t ab ili ty . - Ingic' and electrical schematic diagrans and facilities functional diagrams were presented.

Status:

Closed

34. Using detailed electrical sch ematics and piping ; diagrans, discuss 'the (7.3). automatic i and manual operation and control of the station service ~ water sys tem and the component cooling water sys tem. Discuss the interlo'cks, automatic swi tchover, tes tab ili ty, single' failure ch annel indepe ndence ,

' indication of operability, and the isolation functions.

Response

A detailed discussion using electrical schematics and . piping' diagrams was provided to the staf f. Reference was 'made to FSAR 7.3 and 7.6 regarding pressure switches, FSAR 9.2-1 and 9.2-2 regarding important safety funct ions . Regarding isola tion ~ valve no-go' tes ting , -- the NRC :

requested that a . list be compiled.

Status:

Confirmatory;- pending the . . submittal of an isola tion , val've _ no-go .

test {ng list and a revision to 'FSAR Figure 9.2-4.s m .

35. Ident ify any pneumat ic al ly ope rat ed valves in the ESF system. Us ing (7.3) de t ailed schemat ic s , describe their operation on loss of ins trume nt air system.

Response

There are no pnuematically operated valves in the BVPS-2 ESF sys tems which must operate act ively to ach ieve an ESF safety fun ct io n. The NRC expressed pa rt icular int eres t in the letdown cont aiment iso la-tion valves and the letdown orifice isolation valves . This invo lved a discussion of the cons eque nces of an inad ve rt ant relief to the pres sure relief tank.

S.t a t u_s_:

Closed

36. Discuss the tes t ing provision in the eng i neered safety fe ature P-4 (7.3) interlocks.

Response

The planned test for the P-4 interlocks is a proced ur e involving e nt eri ng the panel to read vo lt ages to verify SI block reset . DIE will condider a design ch ange to ins t al l a t es t ing capability fr om the panel face.

Status:

Open; pending DLC decision on the design change.

37. On May 21, 1981, Wes tinghouse not ified the Commis sion of a potentially (7.3) adverse control and protection system interaction whereby a single random f ailure in the volume control tank (VCT) level control system could lead to a loss of r edundancy in the safety inj ect ion sys ten fo r ce rt ain Westinghouse plant s . Discuss the VCT level control sys t em in Beave r Valley 2 design.

Response

A discussion of a potentially adverse control and prot ect ion sys tem interaction demons trated that a single failur e in VCr level control channel LT-115 involvement , could eventually lead to a los s of redun-dancy in the safety inject ion system. Howeve r, because of the four following bases, this arrar.geme nt is not considered to empr omise plant s a fe t y.

1. Plant operating procedures

2. Charging pump operating characteristics

3. Plant design bases for safety related systems and plant technical specifications

4. Volume control tank (non-safety) design basis i _

Thus, ad e q ua te capabili ty fo r maintaining plant safety was demonstrated.

Status:

j '. Closed

38. Discuss the fault tree analys is (FTA) t echniq ue and the interf ace. with (7.3) WCAP-8760, '.' Failure Mode and E f fect s Analys is of the Engineered - Saf ety i: Features Ac tua tion Sys tem'." Confirm th at the interface requirements speci fied in WCAP-8760 have been met and include a statement in the FSAR

to that ef fect.

I Response:

I i

The interf ace be tween Wes tinghouse (NSSS) and Balance of , Plant (BOP) equipment electric and controls is identified in the Failure Modes and E f fects Analysis (FEA) for the BOP s afe ty sys ten. Each relay.

contac t s ich is part of the Engineered S af e t.y Features Actua tion System is shown in the elementary diagrams of the BOP equipme nt'.

Failure of the relay contact and failure of . the NSSS actuation signal which energizes the ; relay are shown on the fault tree development by

c omponent number ami electrical train designation. his informa tion -

appe ars in the FMEA with the relay actuation signal failure ide nt i-

- fied as the NSSS interf ace. We FEA documents the results of the fault tree analysis.

Status:

, - Confirmatory; pending the. inclusion of a specific s tateme nt in the FSAR - confirming - that all specified interface requirements have been

- met.

39. On August 6, 1982, Wes tinghouse notified the s taf f of a po t ential '

(7.3) unde tectab le failure in online tes t circuitry for the . mas ter relays - in the engineered safeguards systems. We undetectable failure involves. the output (slave) relay continuity proving lamps and their associated shunts provided by tes t pushbut tons . - If .af ter tes ting , a shunt is . not provided

for any proving lamp because of a switch contact

fa ilure , any subsequent safeguards actuation could cause the lamp to' burn open before its associ--

ated slave relay -is energized. his -'would then prevent - actuation of 'any -

as so ci ated safeguards devices on that slave rel ay. Until an accept &le circui t ' modification is designed .' Wes tinghouse has provided tes t' proce-dures - that ensure that the ' slave relay circuits operate normally sen tes ting of the mas ter relays is completed. ' Discuss this ' issue as- applied to Beaver, Valley 2.

Response

%is . item has been reported to the NRC by DLC Significant . Deficiency Report (SDR) 82-04. '

IStatus:

.Open; pending resolution of SDR.82-04.

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40. Verify whether the systems required for safe shutdown can be periodically (7.4) tested during normal ope ra t ion. Provide a cros s-reference to Technical Specification sect ions fo r those compo nent s that wil l be t es t ed during normal operation.

Response

At the ICSB meet ing it was verified that the t es t abili ty of the systems was incorporated into the design. Reference was made to FSAR Sect ion 7.1.2.4 fo r t ho se compo nent s that will not be tes ted during normal operations .

Status:

l Closed  ;

l

41. Use plant design drawings to discuss the main steam power operated relief I (7.4) valve control scheme. Is this a safety grade system? l 1

~

Response: )

A detailed discussion of the control scheme for the main stean atmos-pheric relief valves was provided . This system is a safety gr ade system.

Status:

Closed

42. FSAR Sect ion 7.4.1.2.2 s t ates , " Loss of ins trument air does not pr eve nt (7.4) the operation of the minimum systems necessary for hot standby."

Provide further discussion for valve operat ion in auxiliary fe ed wa t er system, steam generator PORV, RHR sys tem, and other pneumatic operators used in the safe shutdowns systems.

Response

Speci fic valves were uscd to illus trat e that ins trume nt air is not required for any valve to move to its safety position during reactor operation.

Status:

Closed

43. Provide a t ab le showing safe shutdown display infu nnat ion and ident ify safety grade items.

Response

FSAR Table ' 7.5-1 was discussed. The variable designated AI on this t ab le are used to satis fy the safe shutdown dis play information requirements of R.G. 1.97, Rev. 2.

~ - - - _ . - _

.- .- - . -- = _ _

Status:

Closed

44. Describe the capability of achieving hot and cold shutdown frcan outside (7.4) the control room. As a minimuri, provide the following information:

a. location of trans fe r swi tches and remote control statiors (ESP and ASP) (include layout drawings, etc).

b. Design criteria for the remote control station equipment including

- transfer switches.

c. Description of distinct control fe atur es to both res trict and to assure ac ce s s , dien necessary, to the displays and controls loca ted outside the control room,

d. Discuss the testing to be performed during plant operation to verify the capability of maintaining the plant in a safe shutdown condition from outside the control room,

e. Descrip tion of isola tion , separation ami trans fer/ override pr o- ,

vis ions . 'Ih is should include- the design basis- for preventing electrical interaction between the control roan and remote shu tdown  ;

equipment.

f. Description of any ca munication sys tems required to coordinate ,

operator actions, including redundancy and separation.

g. Description of control roan annunciation of remote control or over-ridden status of devices under local control.

h. Means for ensuring that cold shutdown can be accomplished.

i. Discussi the separation arrangement be twe en safe ty-r ela ted and nonsafety-related instrumentation on the auxiliary shutdown panel.

Response

Schematics for the emergency shutdown panel and the ' alternate shut-down panel were pr esented . All of the . items were addres sed and answered satisfactorily.

1 Status:

Closed l

45. Use detailed schematics to describe' the contro1E circuits of the pressur-(7.4) izer pressure control (PORV and heater control), including the interlock ~

and bypass provision from the remote control, panel.

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Response

Elementary diagrams fo r the pressurizer heaters , backup and control groups, were presented and discussed. Refer to Item 16 for the PORV description.

Status:

Closed

46. Discuss the plant ( tes ts to verify the capabili ty of maint aining the

( 7.4 ) plant in a safe shu tdown cond i t ion from ou t s ide the control rom .

Describe design compliance with Regulatory Guide 1.68.2.

Response

Tes ting to verify the capability of maintaining the plant in a safe shutdown condition from outside of the control rom will be done in accordance wi th R.G. 1.68.2. This tes ting is de sc ribed in FSAR Section 14.2.12.6.4 and will be further clarified in the res po ns e to FSAR question 640.03.

Status:

Confirmatory; pending the amendment to FSAR 14.2.12.6.4.

47. Using de t ailed plant design drawings ( sch ema tic s ), discuss the de s ign

( 7.5 ) pe rt aining to bypassed and inoperable s tatu s ind ica tio n . As a minimum, provide the information to describe:

1. Compliance wi th the recommendations of R.G. 1.47. Include a discus-sion of your comments in Section 7.1.2.5 of the FSAR.

2. 'Ih e des ign philosophy used in the selection of equipment /systes to be moni tored . Include a discussion of the logic diagrams in Section 7.5 of the FSAR.

3. How the design of the bypass and inoperable status indication sys tes emply with positions B1 through B6 of ICSB Branch Technical Position No. 21.

The design philosophy should desc ribe as a minimum the cri teria to be employed in the display of inter-relationships and dependencies on equip-ment / systems and should ins ure th a t bypas s ing or delibe ra tely ind uced inoperabili ty of any auxili ary or suppo rt sys tem will au toma tic ally indicate all safety systems af fected.

Response

A discussion was held to describe the bypas sed and inoperable s tatu s indication (BISI) system. The BISI logic di agrams were used to describe and answer questions asked by the NRC reviewer.

Status:

Closed

48. Use schematic and layout drawings to discuss the physical separation (7.5) and wiring for redundant sa fe ty-rela ted ins trument s on the main control board.

Response

An explanation of the methods used to as sur e separation be tween IE and non-instrumentation on the main control board was given.

Status:

Closed

49. Provide a discussion (using de tailed drawings) on the res id ual heat (7.6) removal (RHR) sys tem as it pe rt ains to Branch Technical Positions ICSB 3 and RSB 51 req ui r eme nt s . Spe ci fically, addr es s the fo llowing as a minimum:

a. ne last statement under Section 7.6.2.1 of the FSAR.

b. Testing of the RHR isolation valves as required by Branch Position E.

of BTP RSB 5-1.

c. Capability of operating the RHR from the control roon wi th either onsite or only of fsite power available as required by Position A.3 of BTP RSB 5-1. his should include a discussion of how the RHR sys tem can perform its function assuming a single failure.

d. Describe any operator action required outside the control roan af ter a single failure has occurred and justify.

In addi tion, identify all other points of interface be tween the Reactor Coolant System (RCS) and other systems whose design pressure is les s than that of the RCS. For each such interface, discuss the degree of conform-ance to the requirements of Branch Technical Position ICSB No. 3. Also discuss how the as soci ated interlock circui try confonns to the req uire-ments of IEEE Standard 279. We discussion should include illus tra tions from applicable drawings.

Response

he RCS/RHR interf ace is the only high-pres sure/ low pres sure inter-f ace wh ich falls under the requirement of BTP ICSB 3. FSAR Sections 5.4.7 and 7.6.2 and Figure 5.4-4 were discus sed . Tes t ing of the valves was discussed using FSAR Sec t io n 7.6.2. The capture key method of transferring valve power supplies was discussed in detail .

Status:

Closed

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50. Using detailed system schematics. describe the power distribution for the (7.6) ac cumula tor valves and as soci a ted interlocks and controls including position indica tion in the control room and bypass indica tor ligh t arrangement.

Response

A description of the powe r removal sch eme for the Accumula to r Discharge Isolation Valves was presented.

The NRC reviewer was concerned that indica tion is not provided to indica te s tatu s of the slave cont ac tor and sugge s ted th a t a status ligh t be prov ided . DLC agreed to incorporate a design ch ange to

. address this concern.

Status:

Confirmatory; pending the submission of an acceptable design.

51. Discuss interlocks for RCS pres sure control during low tempe ra ture (7.6) operation.

Response

The interlocks for RCS pressure control during low tenperature opera-tion was discussed in detail and indica ted that the low tanperature overpressure protection system is a safety-grade system.

Status:

Closed

52. Describe the au tomatic and manual design fe a tures pe nni t ting swi tchover (7.6) from the inj ection to the recirculation mode of emergency core cooling ,

including protect ion logic, c omponent bypasses and overrides, parameter monitored and controlled, and test capabilities.

Response

Using electrical sch ema tic s atul logic diagrams, the ' operation of the swi tchover from inj ection to recirculation was described in de tail .

The NRC reviewer expres sed concern ab out the tes tabili ty of the interlock for the charging pump miniflow isolation valves.

Status:

Open; pending the discussion of the - tes ting planned fo r ~ the inter-locks on the above-mentioned valves.

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