ML17333A008

Forwards Revised Section 1.5 to FSAR Re Unresolved Safety Issues.Pages Will Be Incorporated Into Amend 23

ML17333A008
Person / Time
Site:	Columbia
Issue date:	01/11/1982
From:	Bouchey G WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Schwencer A Office of Nuclear Reactor Regulation
References
GO2-82-23, NUDOCS 8201260036
Download: ML17333A008 (57)

Text

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RECIP ~ NAblE'KC!IP>>IEIV>>TIAFFZLIIATIOAi SCHi>>!!ENCKRPA.

Lh censing Branch 2,

SUBJECIT:

Forwa'rds revised Section 1.5 toi FSAR reI unr esoilved saifety issues.Pabes wi1 lt be incor pore'ted intoi A'send 23.,

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Washington Public Power Supply System P.O. Box 968 3000 George Washington Way Richland, Washington 99352 (509) 372-5000 January 11, 1982 G02-82-23 SS-L-02-KSN-82-001

!'20i260036 820iii PDR ADOCK 08000397 A

PDR Docket No. 50-397 Mr. A. Schwencer, Director Licensing Branch No.

2 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Subject:

NUCLEAR PROJECT NO.

2 UNRESOLVED SAFETY ISSUES

'P,EQENt D 9

JAN35]9SV

)

g gm> ttmtte tteuu fgMfStttSX 5 TIOC

.r'efer ence:

Letter, A. Schwencer to R.L. Ferguson, "WNP-2 FSAR - Request for Additional Information", dated November 16, 1981 Enclosed are sixty (60) copies of the revised Section 1,5, Unresolved Safety Issues, per the referenced letter requesti ng additional infor-mation.

These pages will be incor porated into Amendment 23 of the WNP-2 FSAR.

Very truly yours, G.

D. Bouchey Deputy Director, Safety and Security KSN/jca Enclosures cc:

R Auluck -

HRC WS Chin

- BPA R

Feil NRC Site yl gy

~8 c'

r Fl it N

e Fl Pl

~ V

~ cg

~ I-C l~

~

~

1 e-c '"

h

'- MP-2 a % >/%4j,f 5

REQUIREMENTS FOR FURTHER TECHNTCAL XNFORMATZON yI l.

1 DEVELOPMENT OF BWR TECHNOLO. Y

,g

, ~

,1.5.1.@

ACRS Concerns I

This section addiesses those concerns of the Advisor Com-

~ mittee onXReactor Safeguards pertaining to WNP-2. pTable

. 1.5-1 summarizes resolution of ACRS concerns.

'h

~

'.5.1.1.1 Containment Design Features to Minimize Effects of Bypass Leakage The provisions made in the design to minimize bypass leakage (directly from inshde the containment to,outside the reactor building) after a postulated loss-of-coo1ant accident are described in Section

..2.3 of this FS 1.5.1.1.2 Pipe Whip Protec"ion Provisions "The Committee believes:that protection against pipe whip should be provided byithe applic'ant in accordance with criter'a being developed by the AEC Regulatory Staff."

Response

.p'he provisions made in, the design of WNP-2 to provide pro-tection against-dynamia effects'associated with the postu-lated rupture of piping>(pipe wh'p).are described in Section 3.6 of this FSAR.

e 1.5.1.1.3 Design/Criteria for Znactivehumps and Valves

~

~

~

~

~

~

"Active pumps and valves of the reactor coolant pressure boundary required to.perform safety functionshwill be de-signed, to deformation limits for which the calculated primary'tresses will be in the elastic range.

Acceptable design criteria for inactive pumps and valves are yet to~be estab-lished.

<<This matter should, be resolved in a manner satis-facto to the Regulatory Staff."-

Resp nse:

f s item has been resolved and the resolution documented jin e letter of April 23, 1974 from R.C. DeYoung to J.J.

Steep

'8'-2 l 5.l.le4 Main Steam Line Leakage Control Syste".>

'The, applicant has proposed to install a sealing system to ns 'inima leakage through the, main sz,-san'lin '::clat/~:.

valves following a postulated loss-of-coolant accident and has in rogress a s"udy to establ'sh the design of such/a system.

The Committee believes that a sealing system should be instal ed.

This matter should be resolved in a manner satisfactoxg to the Regulatory Staff prior to completion of construction, of the plant."

Response

The provisions ~ade in the design of HNP-2 to m'inimize leak-age through the main steam line isolation valves "ollowing a postulated loss<>of-coolant accident are described in Section 6e7 of this FSAR.

/

1.5.1.1.5 Mitigation of Consequences of Control-Rod Drop Accident s

"Analyses oz postulated, control-rod drop accidents occurring

.in similar cores during certain portions of the fuel cycle indicate unacceptable res&its.

Studies of provisions to reduce the probability of this accident to negligible levels are underway.

This matter spool'd be resolved in a manner satisfactory to the RegulatoryjStaff prior to completion of construction."

ty

Response

The design and procedura.

provisions that are being used on this plant are described in Section~)..3.2.6.

These pro-visions are adequate to control individual rod worths, and ensure the consequences of a postulated rod drop accident are acceptable.

/

l. 5.1. l. 6 Anticipated Transients Without Scram "The applicant has studied design featuresgto make tolerable the consequences of failure to scram during~ynticipated transients, and has concluded that automatic%ripping of the

~

recirculation pumps and injection of boron could provide for a suitable backup 'to the control-rod system for~,this type of event.

The Committee believes that this approach repre-sents a

ubstantial improvement and should be provided for the H

ford No.

2 reactor.

However, further evaluation oz the s fficiency of this approach and the specific means of imp) menting the proposed pump trip should be made.

~This ma&er should be zesolved in a manner satisfactory to "the R gulatory Staff and the'ACRS during construction of thy lant."

1. 5-2

'P

~ ~

~

~

ae 4

onse:

WNP-2 AMENDMENT NO.

13 February 1981 The consequences, of an anticipated transient withou~mscram (AYWS) are mitigated by tripping the recirculation pumps and by manual insertion o~e control rods.

(Fo more information, see 15.8).

'l.5.1.2 Current Development Program 1.5.1.2.1 Loose Parts Detection.>"

A Loose Parts Detection

~S 'em will be provided.

for a system descriptio See 7.7. T.~

WN -2 AMENDMENT NO 14 April 1981

".5.1.2.2 Mark II Containment Suppression Pool Dynamic Loading The Washington Public Power Supply System, in conjunction with other Mark XI. owner utilities, has submitted,a'esign basis document designated as "Mark II Containme~~ynamic Forcing Function Informa tion Report" (DFFIR), NED0-21061, and NEDE-21061P describing the suppressian~pool dynamic loading phenomena during a safety/relief

~v. 'lve actuation or OCA event.

The evaluation of that+Csign basis'ocument against the current lvPPSS Nuclear Project No.

2 design was prepared and submitted to the NRC.~A verification program to demonstrate the'conservatism of thee DFFIR has been sponsored by the Mark II owners" and is describeWin the "Mark II Containment Suppar"ing Program Report",

G Document NED0-21297."

Additiona information concerning plant evaluation,.for suppression pool dynamic loading is contained in the~Plant De~go Assessment Report for SRV and LOCA loads, Reviszoq 2,

~ansmi tted to the NRC as Appendix G to the

FSAR, September 19, 1979.

1.5-4

\\g Ja r ('i 4

A )

~a

~ P A

S Concern

.4, ~

W"IP-2 Resolved*

.Unresolved TABKE 1. 5-1 RESOLU'1'ION OF A~.S

='O."~gE is""

AMENDMENT NO.

13 February 1981 Pr>ge 1

Addressed in FSAR Section (Where A olicable)

ACRS Group I:

1.

NPSg for ECCS X

Pump's 2.

Emergency Power X

'L 3.

Hydrogen~Control X

4.

Instrument Lines X

Penetrating Contain-

'ent 5.

Strong Motion

~

X Seismic Instru-inentat ion 6.

Fuel Pool Design X'g Basis QC'.

Pump Flywheel X jg Missiles 8.

Protection Against X

Industrial Sabotage'.

Vibration Monitor-'

ing I

10.

Inservice Inspec-X tion of RCPB~'1.

Quality Assurance X

During De.sign, Con-struction and Oper-ation 12.

Inspection of BWR X

Steam Lines Beyond Isolation Valves

~13.

ndependent Check-X of Prim'ary System Stress Analysis App'a C g

'R G. 'l.1 p8.3, 7.3.2.1.3 r 6.2.5 7.1.2.4(a) 3.7.4 9.1.2.3, 9.1.3.3 Not Applicable 13.6 1.5.1.2;1, 3.9.2, App.

C, R G.

1.20 5.2.4 1 7 6.6 primary Sys'tom design to ASt)E.

BPV Code Secti n

3.

Section 5.2'.5-5

V

CRS Concern TABLE 1.5-l (Continued)

Resolved*

Unresolved Page 2 o' Addre~ped in FSAR,.S ection (Where~Aoolicable)

Grou I (continued):

14.

Op rational Stabil-X ity f Jet Pumps Confirmed at,

'resden 2 and 3

19.

Diesel Fuel Capacity

~gi 20.

Biological Shield Capability, 21.

Operating One Plant X

while others are under Construction 22.

Seismic Design of X

Steam Lines 23.

Quality Group Classi-

'ication of Pressure Retain'ng Components Ultimate Heat Sink X

X 15.

Pressure Vessel X

Surveillance of Fluence and NDT Shift 16.

Nil Ductility X

Properties of Pressure Vessel X,

Materials 17.

Operation of

+X Reactor with less than all loops in

-service j"-:

t 18.

Criteria for Pre-

>~'

operational Testing

./

gP 5.3.1 5.3.1 Addressed in Millstone 1, Docket No.

50-245 b

~14.1 9.5.4.

3.8.3.4

~ Not Apolicable 3.2.1, 3,.7.2.1.8.2

)

3 2

2 g 3

2 3

1.2.2.12.3, 9.2.'

1.5-6

~ a

~ ~

WNP-2 A:/LED!'fL3(P'O

~

Februa'L y 1981 TABLE 1.5-1 (Continued)

P+ge 3 oi. 7

~

~

A4 S Concern Resolved*

Unresolved Addressed in FSAR~Section (Wnere" A olicable)

Group

( continued):

25.

Ins'rumentation to Detec't,, Stresses in Contairgnent Walls 3.8.2.7, 7.5.1.5 Group IA:

2.

1.

Use of Furnace X

'ensitized stain-less steel Primary system X

detection and location of leaks 3.

Protection against X

pipe whip 4.

Anticipated trans-X ~

,f'ents without scram 5.

~

ECCS capabilities~',

of current and older plants Group IB:

sa 1.

Positive moderator X

coefficient 4

2.

Fixed in-core de-X tectoxrs on high power>~ P WRs 3.

Pefformance of X

critical components

'n post-LOCA environ-ent P

P c'

Ir rt 5.2.3 7.6.1

~ 3 3.6 Will be addressed in response to NASH-1270 6.3.. Conformance to 10 CFR 50, Appendix K addres-sed in 6.3.3 a

Not Aa'plicable 9,

7.5.2, NEDO+10698

l. 5-7

s

WNP-2 t

AMENDMENT NO.

13 February 3.983 TABLE 1.5-1 (Continued)

Page 4 o 7

ACR Concern Addresse in

/

Reso3.ved*

Unresolved FSAR S~tion (Where Aa licable) 6.

Group ZBX,(continued):

4.

Vacuum, relief valves X

controls.ng bypass paths on BWR pressure suppressioh contain-ments 5.

Emergency power. for X

two or more reactors at same site 5

Effluents from light.. X water cooled reactors'~

L 4J~

7.

Control rod ejection X ',s'"

accident g9 Group TC:

ss a

~

g/

1.

Main steam isolation X

valve 'leakage of BWRs 2.

Fuel densification X

2 t

r 3.

Rod sequence control X

system 4.

Seismic>'Category I X

requirements for auxin';iary systems Group TI:

1.

Turbine missiles

,y' X

2 7 Cannainmena Sprays X

a ll~ 3 Not Applicable App.

C)

R.G -

1 ~ 96 s 5.4.5 Topical report NEDM-10735

~7. 7. 1. 11 NEDO-10527 9.1.3, llew 3.1

~ 3, 11 '

3.5.1.3 6.5

~ 2

A

WH: -2 l

EbJ AMENDMENT NO I'>

February 198!

TABLE 1. 5-1 (Con'.- inured',

Pope..

5 o c A~cRS Concern Grouo I (continued):

Resolved*'nresolved Addr>s. e~ in FSAR Seqzion (Where A amicable)

Pressure Vessel Thermal Shock, Post-LOCA

'\\

Xnstrume ts to Detect Fu'el Failure 3.

.5.3.1, 5.3.3.6 7.6.1.1 7.1;2, Appendix Topical Report NEDO-101 89 4.2, 4.3, 4.4,

15. 0, Topical Reports NEDO- : 10174, N DO-10179 NEDO-1 0 20 8, NEDO-10505 and NEDM-10735

..:i'oose Parts X i' 3.9.2g 4.4.6.1 Monitoring 6.

Common Mode

~

X Failures X

j/

J' 8.

'NR Recirculation X

5.4.1.4 Pump Overspeed'uring LOCA ~~

r/

9.

Seismic Scram X

No seismic scram is incorporated.

Seismic instru-mentation meets R.G.

1.12 (App. C).

10.

ECCS capability X

6.3, ',1 5.0 Top-fdr Future Plants ical Reports NED0-10892, NEDO-101'79.

~ 1.5-9.

0

'CN "2

..AENDb)ENT NO.

13 February 1981 CRS Concern Page 6 o Addressed in FSAR Section (Where Aoo~licable)

Resolved*

Unresolved TABLE 1.5-1 (Continu=-d>

RESOLUTION OF ACRS CONCERNS Grou II (continued):

11.

Instrumentation to-Fol&w Course of Accident

. Group IIA:

1.

Pressure 'in'ontain-X ment Followi'ng LOCA E

2.

BWR Control Rod

'",X Drop Accident 3.

Ice Condenser

'ontainments 4.

Rupture of High X

Pressure Lines

'Outside Containment 5.

PWR Pump Overspeed 6.

Isolation of Low X

Pressure from High Pressure Sg'stems A

7.

Steam Generator Tube Leaks

~ 0 8.

ACRS/>NRC Periodic 10 Yeai'eview of Older

~ Reactors X

7.5,;1, 7.5.2 6.2.1.3, Topical Report NEDO-10320, NED0-20345, NEDO-gI 20550 I NEDO 20533 I NEDN-10976, NEDO-1 0 329 15.4.9 Not Applicable 3.6.1, 15.0 Not Applicable V

5.2, 6.2.4.3,

's6

~ 3

~ 2 ~ 2~

7 ~ 3 Not.Applicable L

Not Applicable 1.5-10

Wl/P-2 qQ~

MV~A AMENDMENT NO.

13

February, 1981 fl TABLE 1. 5-1 (Cm>t.l'i;i".,'-

RESOLUTIOR OF ACRS CONCEkhS RS Concern Resolved*

Unresolved Addressees in FSAR Section (Wnere A

licable)

X Grou IIB:

1.

Hybrid Reactor X

Protection Systems 2.

Sx8 BWR;, Fuel X

Qualification Behavior of.<Hark III X

Containment

~4 il'.

Stress Corrosion

, X

, Cracking

/

~

f'roup IIC:

~fl

/~

1.

Locking Out ECCS Power,,

Operated Valves 2.

Fire Protection

~,

X 3.

Design Features to/'

Contxol Sabotage, 1 ~

f 4.

Decontamination; 'and X'ecommissionin'g of Reactors

/f 5.

Vessel Supporting X

Structures

/f S

f'."'.

Water Hammer

/'.

Main enance and In-X spection of Plants

/g 8.,Behavior of Nark I X

Containments

~f/

7.2.;1.1

~

g/

//

/f

/"4. 2

/f/

7.2.2 Not Applicable 3.1.2.4.3, 5.2.3.4, 5.2.4 7.3.1.1.1 8.3.3, 9.5.1 13.6 II Not Applicable 6.3 5.2 3.8 Not

.2.2.5

.4,,3.9.6,

~ 2 o 7'<i Applicable

';,3.9.1, 3.9.3,

'3.9.5

• Resolved means a specified conclusion or policy decision has been reached by the NRC and ACRS.

1. 5-11

i C

I

WNP-2

1. 5 REQUIRENENTS FOR FURTHER TECHNICAL INFORf'lATION 1 '.1 UNRESOLVEO SAFETY ISSUES The NRC staff continuously evaluates'the safety requirenents used in its reviews against new information as it becomes avail-able.

Information related to the safety of nuclear power plants comes from a variety of sources including experience from operating reactorsr research resultsi NRC staff and Advisory Committee on Reactor Safeguards safety reviewsr and vendors architect/engineer and utility design reviews.

Each time a

new concern or safety issue is identified from one or more of these sourcesi the need for immediate action to assure safe operation is assessed.

This assessment includes consideration of the generic implications of the issue-In some casesi immediate action is taken to assure safetyi e.g.i the derating of boiling water reactors as a result of the channel box wear problems in 1975.

In other casesr interim measuresi such as modifications to ooerating proceduresr may be sufficient to allow further study of the issue prior to naking Licensing decisions.

In most cases'owevers the initiaL assessment indicates that immediate Licensing actions or changes in Licen-sing criteria are not necessary.

In any events further study may be deemed appropriate to make judgements as to whether exist" ing NRC staff, requirements should be modified to address the issue for new plants or if backf itting is appropriate for the Long-term operation of pLants already under construction or in operation.

"generic safety issues" icular class or type o

cific plant.

Certain of as "unresolved safety issues".

Resolution of G'eneric Issues dated January 1r 197S.)

issues are considered on a

has made an initiaL deter-nce of the issue does not quire Licensing actions ew is underway.

These issues are sometimes called because they are reLated to a part nuclear facility rather than a

spe t'hese issues have been designated (NUREG-0410'NRC Program for the Related to Nuc Lear Power PLants" i Howevers as discussed above'uch generic basis only after the

.NRC

, mination that the safety signif ica prohib i t cont inued operation or re while the Longer term generic revi 1.5.1.1 ALAB-444 Requirements These Longer term generic studies were the subject of a decision by the Atomic Safety and Licensing Appeal Board of the Nuclear Regulatory Commission.

The decision was issued on November 23'977 (ALAB-444) in connection with the Appeal Board's considera-tion of the Gulf States UtiLity Company application for the River Bend St.ationr Unit Nos.

1 and 2.

1.5-1

klNP-2 This section is specifically included to respond to the decision of the Atomic Safety and Licensing Appeal Board as enunciated in ALAB-444 and as applied to an operating License proceeding involving Virginia Electric and Power Conpany (North Anna Nuclear Power Stations Units 1

and

2) i ALAB-491'RC 245 (1978).

In a related matters as a result of Congressional action on the Nuclear Regulatory Commission budget for fiscal year 1978'he Energy Reorganization Act of 1974 was amended (PL 95-209) on December 13'977 to includei among other thingsi a

new Section 210'Unresolved Safety Issues PLan".'n response to the reporting requirements of the new Section 210'he NRC staff submitted to Congress on January 1i 1978' report describing the NRC generic issues program (NUREG-0410)

~

The NRC program was already in place when PL 95-209 was enacted and is of considerably broader scope than the "UnresoLved Safety Issues Plan" required by Section 210.

In the Letter transmit-ting NUREG-0410 to the Congress on December 30'977'he Commission indicated that "the progress reportsr which are required by Section 210 to be included i'n future NRC annual reportsi may be more useful to C'ongress if they focus on the specific Section 210 safety items.

In 1978'he NRC undertook a review of over 130 generic issues addressed in the NRC progran to determine which cases fit this description and qualify as "Unresolved Safety Issues" for reporting to the Congress.

The NRC review included the deveLopment of proposals by the NRC staff and review the final approvaL by the NRC Commissioners.

This review is described in a reports NUREG-0510'ntitled "Identification of Unresolved Safety Issues Relating to Nuclear Power Plants A Report to Congress"i dahed January 1979.

The report provides the folLowing definition of an "Unresolved Safety Issue".

"An Unresolved Safety Issue is a matter affecting a number of nuclear power plants that poses impor-tant questions concerning the adequacy of existing safety requirements for which the finaL resolution has not yet been developed and that invoLves con-ditions not Likely to be acceptable over the Life-time of the plants it affects."

Further'he report indicates that in appLying this definitions matters that pose "important questions concerning the adequacy of exis'ting safety requirements" were judged to be those for which resolution is necessary to (1) compensate for a possible major reduction in the degree of protection of the pub'Lic health and safetyr or (2) provide a potentially significant 1.5-2

'WNP-Z'ecrease in the risk to the public health and safety.

Quite simplyi an "Unresolved Safety Issue" is potentially sionifi-cant from a public safety standpoint and its resolution is LikeLy to result in NRC action on the affected plants.

All of the issues addressed in the NRC program were systemati-caLly evaluated against this definition as described in NUREG-0510.

As a results 17 "Unresolved Safety Issues" addressed by 22 tasks in the NRC program were identified.

The issues are Listed below.

Progress on these issues was first discussed in the 1978 NRC Annual Report.

The number(s) of the generic task(s)

(e.g.i A-1) in the NRC program addressing each issue is indicated -in parentheses following the title.

"UNRESOLVED SAFETY ISSUES" (APPLICABLE TASi< NOS.)

1.

2.

3.

4 5.

6.

7.

8

~

9.

10.

11.

12.

13.

14.

15.

16.

17.

Water Hammer (A-1)

Asymmetric Blowdown Loads on the Reactor Coolant System (A-2)

Pressurized Mater Reactor Steam Generator Tube integrity (A

3w A 4i A 5)

BWR Nark I and Nark II Pressure Suppression Containments (A 6i A-7r A-8r A 39)

Anticipated Transients Without'cram - (A-9)

BMR Nozzle Cracking (A 10)

Reactor Vessel Nateria ls Toughness

- (A-11)

Fracture Toughness of Steam Generator and Reactor CooLant Pump Supports (A-12)

Systems Interaction in Nuclear Power Plants (A-17)

Environmental Qualification of Safety-Related ElectricaL Equi pment (A-24)

Reactor Vessel Pressure Transient Protection (A-26)

Residual Heat RemovaL Requirements (A-31)

Control of Heavy Loads Near Spent Fuel (A 36)

Seismic Design Criteria " (A-40)

Pipe Cracks at Boiling Water Reactors

" (A-42)

Containment Emergency Sump Reliability (A-43)

Station BLackout " (A-44)

Six of the 22 tasks identified as the "Unresolved Safety are not appL icable to WNP-2 because they appLy to pressur water reactors only.

These tasks are A-2r A3i A4i A-5i and A-26.

Also'ask A-6 and A-7 only apply to Nark I bo water reactor containments.

The NRC staff has issued NUR reports providing its proposed resolution of the seven '.of remaining tasks that are applicable to WNP-2.

Below is a

of those issues.

Issues" ized A-12'ling EG 14List 1.5-3

Wi II h

fi I

'4

WNP-2 Task No.

NUREG Report and Title Addressed in FSAR Section A-8 NUREG-0487'Nark I,I Containment Lead Plant Pr ogram Load Eva Lu-ation and Acceptance Criteria"r October 1978.

SuppLement 1

to NUREG-0487'ctober 1980.

Supple-ment 2 to NUREG-0487'ebruary 1981.

6.2i Appendix G

A-10 A-24 NUREG-0619m "BWR Feedwa ter Nozzle and Control Rod Drive Return Line Nozzle Cracking" NUREG-0588r "Interim Staff Posi-tion on Environmental Qualifica-tion of Safety-Related ELectrical Equipment" 5.2 3.11 A"31 SRP 5.4.7 and BTP 5-1r "Residual Heat Removal Systems" incorporate requirements of USI A-31.

5.4 A-36 NUREG-0612'Control of Heavy Loads

,9.1 at Nuclear Power Plants".

A-39 NUREG-0487 and SuppLement 1 to NUREG-0487 (see above).

6.2i Appendix G

A-42 iVUREG-0313'evision 1i "Tech-nicaL Report on NateriaL Selec-tion and Processing Guidelines for BWR Coolant Pressure Boundary Piping".

5.2 The remaining issues applicable to WNP-2 are Listed below:

1.

2.

3.

4.

5.

6.

7.

'Mater Hammer (A-1)

Anticipated Transients Without Scram (A-9)

Reactor Vessel Materials Toughness (A-11)

Systems Interaction in Nuclear Power PLants (A-17)

Seismic Design Criteria (A-40)

Containment Emergency Sump Reliability (A-43)

S.tation B lackout (A-44)

The applicability and bases for Licensing prior to ultimate resolution of the above Listed Unresolved Safety Issues are discussed in Section 1.5.2.

1.5-4

WNP-2 1.5.1.2 New "Unresolved Safety Issues" An in-depth and systematic review of generic safety concerns identified since January 1979 has been performed by the NRC staffs to determine if any of these issues should be designated as new "Unresolved Safety Issues".

The candidate issues originated from concerns identified in NUREG-0660'NRC Action Plan as a Result of the TNI-2 Accident" r ACRS recommendationsr abnormal occurrence reports and other operating experience.

The staf f '

proposed L i st was reviewed and comnented on by the ACRSr the Office of Analysis and EvaLuation of Operational Data (AEOD) and the Off ice of Policy. Eyaluat ion.

The ACRS and AEOD also proposed that severaL additionaL "Unresolved Safety Issues" be considered by the Commission.

The Commission considered the above information and approved the following four new Unresolved Safety Issues":

A"45 Shutdown Decay Heat Removal Requir ements A-46 Seismic Qualification of Equipment in Operating PLants A,.47 Safety Implication of Control Systems A-48 Hydrogen Control measures and Effects of Hydrogen Burns on Safety Equipment A description of the above process together with a List of the issues considered is presented in NUREG-0705'Identification of New Unresolved Safety Issues Relating to Nuclear Power PLantsi Special Report to Congress" i dated Harch 1981.

An expanded

'discussion of each of the new "Unr esolved Safety Issues" is also contained in NUREG-0705.

The appLicabiLity and bases for licensing prior to ultimate resolut ion of the four new Unresolved Safety Issues for WNP-2 are also discussed in Section 1.5.2.

1.5.2 DISCUSSION OF UNRESOLVED SAFETY ISSUES AS THEY RELATE TO WNP"2 A-1 Water Hammer Description.

Water Hammer events are intense pressure pulses in fluid systems caused by any one of a

number of mechanisms and system conditions such as rapid condensation of steam pockets'team-driven slugs of wateri pump startup with partially empty Linesr and rapid 1.5-5

I

WNP-2 valve motion.

Since 1971 there have been over 200 incidents invoLving water hammers in BMRs and PMRs reported.

The water hammers (or steam hammers) have involved steam.generator feed rings and pipingi the RHR systems ECCS systems and containment sprays service wateri feedwater and steam Lines.

Nost of the damage reported has been relatively minors involving pipe hangers and restraints; howevers there have been several incidents which have resulted in piping and valve damages.

MNP-2 P os i t i on:

WNP-2 has installed a

system to preclude water hamner f rom occurring in emergency core cooling system lines.

This system consists of water Leg pumps to keep the ECCS Lines water-filled so that ECCS pumps wiLL not start pumping into voided Lines and steam wiLL 'not coLLect in the ECCS piping.

To ensure that the ECCS Lines remain water"fiLLedr vents have been installed and a technical specification requirement to periodicaLLy vent air from the Lines has been imposed.

Approaches used at design stage include:

(1) increasing valve closure timesr (2) piping Layout to preclude water slugs in steam Lines and vapor formation in water Linesr (3) use of snubbers and pipe hangersi and (4) use of vents arid drains.

MNP-2 has committed to conduct a preoperational vibration dynamic effects program in accordance with Section III of ASNE for aLL Class 1

and CLass 2 piping systems and piping restraints during startup and initial operat.ion.

These tests wiLL provide adequate assurance that, the piping and piping restraints have been designed to withstand dynamic ef fects due to valve c Losuresi punp t r i ps ard other opera t ing modes associated with the'esign operat iona L transients.

Nonetheless~

in the unlikely event that a

Large pipe break did result from a severe water hammer events core cooling is assured by the emergency core cooling systems and protection against the dynamic effects of such pipe breaks inside and outside of containment is provided.

In the event that Task A-1 identifies potentially significant water hammer scenarios which have not explicitly been accounted for in the design and operation of MNP-2r corrective measures may be required at that time.

The task.

has not, identified the need for measures beyond those already implemented.

Based on the foregoingi we conclude that MNP-2 can be operated prior to ultimate resolution o

this generic issue without undue risk to the health and safety of the public.

1.5-6

'WNP

-2'-9 Antici ated Transients Without Scram

==

Description:==

Nuclear plants have sa consequences of tempor "anticipated transient ting conditions may be may impose significant anticipated transients reaction (initiating a

the generation of heat safety measure.

If th cipated transient" and "scram" as desiredi th scram"i or ATWSi would fety and control systens to Limit the a'ry abnormaL operating conditions or s".

Some deviations from normal opera-,.

minorr othersr occurring less frequentlyr demands on plant equipmert.

In some rapidly shutting down the nuclear "scram")r and thus rapidly reducing in the reactor corer is an important ere were a potentiaLLy severe "anti-the reactor shutdown system did not en an "anticipated transient without have occurred.

WNP"2 Position:

A recirculation pump trip provision has been incorporated in the WNP-2 design.

In additionr the Suppl,y System has implemented emergency procedures and operator training to cope with potential anticipated transient without scram events.

Operator training and action as desiredr in conjunct the automatic recirculation pump trip significantly the capability, of the facility to withstand a range anticipated tra'nsient without scram eventsi such tha tion of this facility presents no undue r'sk to the and safety of the public while this matter is under review.

The ATWS issue is currently being reviewed Commission.

A proposed rule was published in the Fe Register on November 24'981.

This proposed rule i

presently being reviewed by the Supply System.

The System wiLL be required to meet the requirements of ATWS rulei which is anticipated in mid-to-Late 1982.

ion with inprove oft opera-health NRC by

". he deraL s

Supply the f inaL A-11 Reactor Vessel Naterial Tou hness

==

Description:==

Because the possibiLity of failure vessels designed to the ASi'lE Boile is remoter the design of nuc lear f protection against reactor vesseL accum'u late more and more service t

reduces the material fracture toug margin.

of nuclear reactor pressure r

and Pressure Vessel Code acilities does not pr ovide failure.

Howevers as plants imer neutron irrad iat ion hness and initial safety

1. 5-7

L'i 1

MNP-2 Results from reactor vesseL surveillance programs indicate that up to approximately 20 operating nuclear reactors will have beltline materials with marginal toughnessr relative to the requirements of Appendix G and H of 10CFR50 after comoaratively short periods of operation.

For most plants now in Licensing process'urrent criteria'ogether with the materials currently employedi are adequate to ensure suitable safety margins for reactor vessels throughout their design Lives.

Howevers a

few plants under Licensing reviews have reactor vessels that have been identified as having the potentiaL for marginaL fracture, toughness within their design Lives; these vessels will. have to be reevalu'ated in the Light of the new criteria for Long-term acceptability.

WNP-2 Position:

The materials used in fabricating the WNP-2 vessel were seLected to assure that suitable safety margins wiLL exist over the Life of the plant; including the degrading effects of radiation on materiaL toughness.

Additionallyr MNP-2 reactor will be operated with restrictions imposed by technical specificat ions on the pressure during heatup and cooLdown operations.

These restrictions assure that the reactor vesseL wiLL not be subjected to a combination of pressure and tempera-ture that could cause britt le fracture of the vesseL if there were significant flaws in the vessel material.

The effect of neutron radiation on the fracture toughness of the vessel material over the Life of the plant is accou'nted for in technicaL specification Limitations.

Based on the informat ion included in FSAR Sect ion 5.3r we can conclude that MNP-2 wiLL have adequate safety margins against brittle failure during operatingr testings maintenancer and anticipated transient conditions over the Life of the plant.

Naterial surveiLLance programs (FSAR Section 5.3.1) and Inservice Inspection Programs (FSAR Section 5.2.4) are in accordance with applicable ASEN Code requirementsi and provide assurance that brittLe fracture control and pressure vesseL integrity will be maintained throughout the service lifetime of the reactor pressure vessel.

Based on the foregoingr we conclude that MNP-2 can be operated prior to ultimate resolution of this generic issue without undue risk to the health and safety of the public.

1. 5-8

f

MNP-2 A17 S

stems Interaction in Nuclear Power Plants

==

Description:==

The design and analyses by the plant designersi and the subsequent review and evaluation by the NRC staff take into consideration the interdisciplinary areas of concern and account for systems interaction to a

Large extent.

Furthermorei many of the NRC regulatory criteria are aimed at controlling the risks fron systens interactions.

Examples include the single failure criterion and separation criteria.

Neverthelessr there is some question regarding the interaction of various plant systemsr both as to the supporting roles such systems play and as to the effect one system can have on other systemsi particularly with regard to whether actions or con-sequences could adversely affect the presuned redundancy and independence of safety systems.

The problem to be resolved by this task is to identify where the present designs analysis'nd review procedures may not acceptably account for potentially adverse systems interaction and to recommend the regulatory action that should be taken to rectify deficiencies in the procedures.

(Also'ee T!!I Action Plan NUREG-0660'tem II.C.3.)

WNP"2 Position:

MNP-2 has been designed to meet the Licensing requirements such as physicaL separation and independence of redundant safety systems and protection against events such as high energy Line rupturesi missilesi high <<indsi floodingi seismic eventsi fires'perator error sr and sabotage.

The design pro'visions are supplemented by NRC staff review procedures of the Standard Review Plan which require interdisciplinary reviews of potentiaL systems interactionsi provide

'.or an adequately safe situation with respect to such interactions.

The quality assurance program which is followed during designs cons'tructioni and operationaL phases for MNP 2 is expected to provide added assurance against the potential for adverse systems interactions.

In mid-1977'ask A-17 was initiated by NRC to confirm the present review procedures and safety, criteria provide an acceptabLe Level of redundancy and independence for systems required for safety by eva luating the potentiaL for undesirable interactions between and among systems.

The NRC staff's current review procedures assign primary respon-sibility for review of various technicaL areas and safety systems to specific organizationaL units and assign secondary responsibiLity 1.5-9

n

to other units where there is a -unctionaL or interdisciplinary relationship.

The Supply System foLLowed somewhat similar procedures and orovided for interdisciplinary reviews and analyses of systems.

Task A-17 provided an independent study of methods that could identify inportant systems interactions adverse y impacting safety; and which are not considered by current 'review procedures.

The first phase of this study began in Nay 1978 and was comoleted in February 1980 by Sandia Laboratories under contract to the NRC staff.

The Systems Interaction Reactor Regulation in A

art,methods that can oe The initial efforts sup is underway; a

range of for feasibility against candidates derived from Branchy formed in the Off ice of Nuc Lear priL 1980'as been studying state-of-the-used to predict systems interactions.

ported by three laboratory contractorsi methods is being considered and tested a

sample of some systens interaction Licensee Event Report evaLuations.

It is expected that the development of systematic ways to identify and evaluate systens interactions will reduce the Likelihood of common cause failures resulting in the loss of plant, safety functions.

Howevers the studies to date indicate

'that current review procedures and criteria supplemented by the application o-post-TNI findings and risk studies provide reasonable assurance that the effects of potentiaL systems interaction on plant safety wiLL be within the effects on plant safety previousLy evaluated.

Thereforer we conclude that there is reasonable assurance that MNP-2 can be operated prior to the fina l resolution of this generic issue

~without endangering the health and safety of the public.

A 40 Seismic Desi n Criteria Shor

-Term Pro ram

==

Description:==

NRC regulations require that nuclear power plant structuresi sy'stems'nd components important to safety be designed to withstand the effects of naturaL phenomena such as earthquakes.

Detailed requirements'nd guidance regarding the seismic design of nuclear plants are provided in the NRC regulations and in regulatory guides issued by the Commission.

Howevers there are a

number of plants with construction permits and operating Licenses issued before the NRC's current regulations and regulatory guidance were in place.

For this reasons re-reviews of the seisnic design of various plants are being undertaken to a,ssure that these plants do not present an undue

I

~

I

WNP-2 risk to the pub l ic.

Task A-40 i sr in e f feet i a

compendium of short-term efforts to support'uch reevaluation efforts of the NRC staffs especiaLLy those related to older operating plants.

In additionr some revisions to sections of the Standard Review Plan and regulatory guides to bring them more in Line with the state-of-the-art will result.

WNP-2 Position:

WNP-2 plant structuresi systems'nd components important to safety are designed to withstand the effects of naturaL phenomena such as earthquakes using current Licensing criteria and requirements.

c The seismic design basis and seismic design of WflP-2 are detailed in FSAR Sections 3.7 and 3.8.

Should the resolution of Task A-40 indicate a

change is needed in these Licensing requirementsi WNP-2 design wiLL be evaluated accordingly.

Thereforei we conclude WNP-2 can be operated, prior to uLtimate resoLution of this issue without endangering the health and safety of the public.

A-43 Containment Emer enc Sump Performance

==

Description:==

FoLLowing a postulated in the reactor coolant the break would be coL water would be recircu the emergency core coo This water may also be spray system to remove drywell and wetweLL at water from the suppres cooling and containmen Loss-of-coolant accidentr i.e.i a break system pipingr the water flowing from Lected in the suppression pool.

This Lated through the reactor system by Ling pumps to maintain core cooling.

circulated through the containment heat and fission products from the mosphere.

Loss of the abiLity to draw sion pool could disable the emergency t spray system.

Concern addressed in this issue ior boiling water reactors is Limited to the potential for degraded emergency core cooling system performance as a r'esuLt of thermal insulation debris that may be blown into the suppression pooL during a Loss-of-coolant accident and cause blockage of the pump suction Lines.

WNP-2 Position:

The blown off insulation panels constitute the only credible debris within the primary containment following a

LOCA and seismic event.

Large pieces of debris are not considered to 1

% 5 1 1

n f

'aJNP-2'ave.

deleterious effects on the containment systems.

The grating at the 501'-0" elevationr which covers approximately 80K of the primary containment cross-sectional area wouLd stop the majority of the Loose insulation panels.

The potential debris in the drywelL could only be swept into the suppression pool via the downcomer piping.

However~

the downcomer pipes (aoproximateLy two feet in diameter) are capped with jet deflectors and would prevent Large pieces fron reaching the suppression pool.

In additions each FCCS pump suppression pooL suction consists of a pipe 'T'ith a

suction screen assembLy at each end.

Accordinglyr we conclude that WNP-2 can be operated orior to ultimate resolution of this generic issue without endangering the heaLth and safety of the public.

A-44 Station Blackout

==

Description:==

The unlikelyr but possibler Loss of aLL AC power (that isi the Loss of AC power from the offsite source and from the onsite source) is referred to as a station blackout.

In the event of a station blackouts the capabiLity to cool the reactor core would be dependent on t'e availability of systems which do not require AC power suppliesr and on the ability to restore AC power in a timeLy manner.

The concern is that the occurrence of a station blackout may be a relatively high probability event and that the consequences of this event may be unacceptablei for examples severe core damage may result.

Review your plant operations to determine your capability to mitigate a station blackout event and promptly implement as necessaryr emergency procedures and a training program for station blackout events (for details see NRC Generic Letter 81-04m dated February 25'981).

MNP-2 Position:

The Loss of all alternating current power was not a design basis event for MNP-2.

Nonethelessi a combination of designr operatingi and testing requirements implemented by the Supply System will assure that WNP-2 wilL have substantiaL resistance to a

Loss of all AC power and that~

even if a loss of a ll AC power should occurs there is reasonable assurance

'that the core will be cooled'hese are described below.

'1.5-12

k f

t

't l

H

WNP-2 gc) LA+(l A Loss of offsite AC power involves a

Loss of both preferred and backup sources of offsite power.

Loss of aLL offsite power for WNP-2 is a relatively unl 'kely event.

The plant is tied into the BonneviL'Le Power Grid which.is based mostly on hydroelectric sources and, is considered one of the most reLiable major distri-bution systems'ased on ac data.

The designi inspectioni and testing provisions for the offsi.e power system are desc'ribed in Section 8.2 of the FSAR.

Data c ited in a recent report (Reference 1.5-1) indicate that the mean time between failures for simuLtaneous Loss of both. independent Lines in the general Location of WNP-2 is about 30 years for outages between two-second and twenty-minute durationr and it is over 110 years for outages over twenty-minute duration.

These figures are based on Bonreville Power Administration rec'ords.

If offsite AC is Losti three diesel generators (two emergency and one HPCS) and their associated distribution system will.

deliver emergency power to safety-related equipment.

The design~ testingr surveiLLancer and maintenance provisions for onsite emergency diesels are described in Section 8.3 of the FSAR.

If both offsite and onsite AC power are Losti WNP-2 may use a

combination of safety/relief valves and reactor core isolation system to remove core decay heat without reliance on AC power.

These systems assure that adequate cooling can be maintained for at Least two hours which allows time for restoration of AC power from either offsite or onsite sources.

The issue of station bLackout was also considered by the Atomic Safety and Licensing Board (ALAB-603) for the St.

Lucie No.

2 facility.

In additioni in view of the completion schedule for Task A44i the Appeal Board recommended that the Commission take expeditious action to ensure that other plants and their opera-tors are equipped to accommodate a station blackout event.

Consequentlyr NRC requested (Generic Letter 81-04'ated February 25'981) a review of plant operation to determine the Supply System's capability to mitigate a station blackout event and promptly implementr as necessaryr emergency procedures'nd training programs for station blackout events.

Appropriate review of procedures and training programs for station blackout events wiLl be completed prior to fueL Load date.

Based on the above'e conclude that there is reasorable assurance that WNP-2 can be operated prior to the ultimate resolution of this generic issue witho'ut endangering the health and safety o'f the public.

1.5-13

0 1

1 n

WNP-2 A-45 Shutdown Deca Heat Removal Re uirements

==

Description:==

Following a reactor shutdowns the radioactivity decay of fission products continues to produce heat (decay heat) which must be removed from the primary system.

The principaL means for removing this heat i'n a boi ling water reactor while at high pressure is via the steam Lines to the turbine condenser.

The condensate is normally returned to the reactor vessel by the feedwater systems howevers the steam turbine-driven reactor core isolation cooling system is provided to maintain primary system inventoryi if alternating current power is not available.

When the system is at Low pressurei the decay heat is removed by the" residual hea't removal systems.

This "Unresolved Safety Issue" will evaluate the benefit of oroviding alternate means of decay heat removal which could substantially increase the plants'apabiLity to handle a broader spectrum of transients and accidents

~

The study will consist of a

generic system evaluation and wiLL re.su lt in recommendations regarding the desirability ofr and possibLe design requirements fore improvements in existing systens or an alternating decay heat removaL nethod if the improvements or alternative can significantly reduce the overall risk to the public.

WNP-2 Positio'n:

The WNP 2 reactor has various methods for the removaL of decay heat.

After a reactor tripi three "regular" modes of ooeration at high system pressure are available for heat removal and coolant makeup:

1.

If turbine bypass'ain condenser and feedwater pumps are availabler one of the feedpumps is used for coolant makeup.

Steam is released through the four main steam Lines and the turbine bypass into the main condenser.

The turbine bypass opens periodically on a pressure signal such that no relief valve actuation is required.

4 2.

If any of the above needed equipment is not available~

or if containment isolation has occurredr the HPCS (high pressure core spray) system together with the RCIC (reactor core, isolation cooling) system is used for about 25 minutesi after which the RCIC system alone is sufficient.

in case the HPCS system is not availabler the RCIC system can be used alo'ne.

In that caser some Level decrease in the reactor vessel wiLL occur during the first 25 minutesi and some steam wiLL be released through the re lief valves during tha-t time.

The reactor will operate at a saturation 1.5-14

hi

,~

MNP -?

pressure of 1r076 psigi which is the setpoint of the two Lowest pressure relief valves.

Condensation of the released stean wiLL result in a nodest heatup of the suppression pool.

3.

If it is desired to avoid reactor cooldown (maintain hot shutdown conditions)i the RCIC is used together with the RHR (residuaL heat removaL) system in the steam condensing mode.

Reactor stean is routed through pressure reducing valves to the RHR heat exchangers where it is condensed.

The condensater together with the RCIC turbine condensatei is pumped back into the. reactor by the RCIC injection pump.

Several variations of these operational modes can be envisioned.

One additionaL

.source of high pressure coolant are the control rod drive pumps.

On older reactor designsr which have a

steam driven HPCI (high pressure coLLant injection) system instead of an electricaLLy driven HPCS systems there have been cases where the HPCI system was not immediately avaiLable on demand.

The RCIC system then was us'ed alone<

and in alL cases known to usi the HPCI system was brought on Line within about 30 minutes.

In the unlikely case that none of the above mentioned equipment can be operatedi manuaL or automatic depressurization of the reactor is available.

Depressurization will decrease the vessel pressure to below 300 psig.

where LPCS (Low pressure core spray) and LPCI (low pressure coolant injection) can be used.

The RCIC and the HPCS at MNP-2 have improvements over comparable systems at older boiling water reactors.

The reactor core isolation cooLing system has been upgraded to safety-grade quality (now required for all boiling water reactors)i and the high pressure core spray is powered by its own dedicated dieseL so it can operate with an assumed loss of alL other sources of alternating current power.

Also',the residuaL heat removal system contains three pumps; the fLow capacity of any single punp is sufficient to easily rem'ove the decay heat.

AccordingLyr we concLude that WNP-2 can be operated prior to ultimate resolution of this generic issue without endangering the health and safety of the publ ic.

A-46 Seismic Qualification of E uioment in 0 eratin Plants Description

'he design criteria and methods for the seismic qualification of mechanicaL and electricaL equipment in nuclear power plants have undergone significant change during the course o>

the commerciaL 1.5-15

ll

nucl,ear power program.

Cons'equentlyi the margins of safety provided in ex isting equipment to resist seismically induced loads and perform the intended safety functions may vary considerably.

The seismic qualification of the equipment in operating pLants musty thereforer be reassessed to ensure the ability to bring the plant to a safe shutdown condition when subject to a seismic event.

The objective of this Unresolved Safety Issue is to establish an explicit set of guidelines that could be used to judge the adequacy of the seismic qualification of mechanicaL and electrical equipment at all operating plants in Lieu of attempting to backfit current design criteria for new plants.

This guidance wiLL concern equipment required to safeLy shut down the plant~

as well as equipment whose funtion is not required for safe shutdowns but whose failure could result in adverse conditions which might impair shutdown functions.

WNP-2 Position:

WNP-2 electricaL and mechanical equipment are designed using seismic criteria delineated in IEEE 344-1971 and the design is being reviewed and approved by the NRC staff using current design criteria and methods for seismic quaLification.

A-47 Safet Implications of Control S

stems

==

Description:==

This issue concerns the potential for accidents or transients being made more severe as a result of control system failures or malfunctions.

These failures or malfunctions may occur independently or as a result of the accident or transient under consideration and wouLd be in addition to any control system failure that may have initiated the event.

ALthough it is generally believed that controL system failures are not Likely to resuLt in Loss of safety functions which could lead to serious events or result in conditions that safety systems are not able to cope withi in-depth studies have not been performed to support this belief.

The potential for an accident that would affect a

particular control system -- and the effects of the control system failures -- will differ'rom plant to pLant.

Thereforei it is not Likely that it will be possible to develop generic answers to these concernsr but rather plant-specific reviews wiLL be required.

The purpose of this Unresolved Safety Issue is to define generic criteria that may be used for plant-specific reviews.

A specific subta sk of this issue will be to study.the steam generator overfilL transient in PWRs and the reactor overfiLL transient in 8WRs to determine and define the need for preventive and/or mitigating design measures to accommodate this transient.

1.5-16

F t

f

MNP-2 WNP-2 P os i t i on:

The Wb'P-2 control and safety systems have been designed with the goal of ensuring that control system failures (either single or muLtiple failures) will not prevent automatic or manual initiation and operation of any safety system equipment required to trip the plant or to maintain the plant in a

sa e

shutdown condition following any "anticipated operationaL occurrence" or "accident".

This has been accomplished by either providing independence between safety and nonsafety systems or providing i'solating devices between safety and nonsafety systems.

These devices preclude the propagation of nonsafety system equipment faults such that operation of the "safety system equip-ment is not impair ed.

A wide range of bounding transients and accidents is present ly analyzed to assure that the postulat d events would be adequately mitigated by the safety systems.

In add',tioni in response to NRC Questions 031.135'31.137'nd 031.138'ystematic reviews of safety systems is being performed with the goal of ensuring that the control systen faiLures (single or multiple) will not defeat safety system action.

Specificallyr these reviews will include:

1.

IE BuLLetin 79-27 (Question 031.135)

A series of tables wiLL be developed which L

sources to incLude alarm indicationsr instru controL devices on these power sources.

The secondary effects on safe shutdown from Loss sources to each Load wiLL be analyzed.

Desi procedure modification will be made as neces determined effects have an adverse impact on i st s power ments and primary and of the power gn and/or sary when the plant safety.

2.

Control System Failure (Question 031.138)

The review procedur e being folLowed to address this question is' 0

b.

C ~

e.

Define bus structure throughout the plant from the grid to the Lower voltage Level.

Identify loads-Eliminate Loads where failure or malfunction would not impac t plant sa fety.

Identify instruments (related to the above non-eliminated Loads) on common instrument taps/impulse Lines/hydraulics.

List effects of failure or malfunction of each Load/

common sensors due to failure or malfunction of power source/pLugged or broken L ines.

1.5.17

MNP-2 f.

Analyze the effects on systems important to plant safety.

g.

Analyze the combined effects due to plugged or broken L ines/cascading power Losses.

h.

Compare results of these analyses to those already covered in Chapter 15.

i.. Nodify and/or augment Chapter 15 as necessary such that failures or malfunctions of common power source or sensor would not require action or response beyond

-the capability of operations or safety systems.

3.

IE Information Notice 79-22 (Question 031.'137)

A matrix will be developed to indicate the effects, of non-safety grade/controL equipmenti subjected to the adv'erse environment of a high energy Line breaks or the protection functions performed by the safety grade equipment.

If interaction is discovered then the impact of failure of the applicable system upon the safety anaLyses wiLL be evaluated.

Several early boiling water reactors have experienced rea vesseL overfiLL transients with subsequent two-phase or L

flow through the safety/reLief valves.

Following these e

eventsr high level trips (Level 8) have been installed at and other operating BIJRs to terminate f Low from the appro systems.

These high-Level trips are single faiLure proof periodic surveillance is required by the Technical Specif No overfiLLing events have been reported at this or other Bl)Rs since the Level 8 trips were installed.

ctor iquid arly

'NP-2 priate and ications.

operating Based on the above'he Supply System concludes that there is a

reasonabLe assurance that MNP-2 can be operated prior to the ultimate resolution of this generic issue without endangering the health and safety of the public.

A-48 H dro en ControL Neasures and Ef fects of H dro en Burns on Sa f e t Equi pment

==

Description:==

Postulated reactor accidents which result in a degraded or melted core can result in generation and release to the containment of Large quantities of hydrogen.

The hydrogen is formed from the reaction of the zirconium fuel cladding with steam at high tempera-tures and/or by radioLysis of water.

Experience gained from the Tf1I-2 accident indicates that we may want to require more specific 1.5.18

WNP-2 design provisions for handling Larger hydrogen releases than currently required by the regulations particular ly for sma L Lerr Low pressur e containment designs.

This issue will investigate means to predict the quantity: and release rate of hydrogen following degraded core accidents and various means to cope with Large releases to the conta',nment such as inerting of the containment or controlled burning.

The potential effects of proposed hydrogen controL measures on safety including the ef fee'ts of hydrogen burnes on safety related equip-ment wiLL al so be invest igate.

WNP-2 Position:

The accident at TMI-2 o'n March 28'979 resuLted in hydrogen generation weLL in excess of the amounts specified in 10 CFR Section 50.44.

As a resuLt of this knowledge it became apparent to NRC that specific design measures are needed for hand ling Larger hydrogen releasesr particularly for smal.leri Low-pressure containments.

As a results the Comnission determined that a

ruLemaking proceeding should be undertaken to define the manner and extent to which hydrogen evolution and other effects of a

degraded core need to be taken into account in plant design.

An advance notice of this rulemaking proceeding on degraded core issues was published in the Federal Re ister on October 2i 1980.

Recognizing that a

n this rulemak ing proc relat ive to hydrogen nented.

These inter October 2i 1980 rede containments (Mark specified that inert umber of years may be required to complete eedingi a set of short-term or interim actions controL requirements was developed and inple-im measures were described in a

second ral

~Re i ster notice.

For plants with seal l and Mark II) su'ch as WNP-2r the interim ruLe ing is required to preclude hydrogen burning.

WNP-2 has committed to inerting the containment buildings during power operation.

The Supply System concludes that WNP-2 can be operated prior to resolution of this unresolved safety issue and the proposed rulemaking without undue risk to the health and safety of the public.

1.5.19

C

~ S L+

1.5

References:

1.5-1 HPPSS-ENT-087'eliability Analysis of the Auxiliary Heat Removal Systems for the Washington Nuclear Project Number 2i April 1981'ashington Public PoMerr Richlandi Washington.

1.15-20

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