ML17194A357

Forwards Containment Sys Branch Draft Evaluation of SEP Topic VI-4, Containment Isolation Sys. Defined Basis Upon Which Specific Configurations Were Judged Acceptable Are Requested as Part of Comments on Rept

ML17194A357
Person / Time
Site:	Dresden
Issue date:	12/18/1981
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Delgeorge L COMMONWEALTH EDISON CO.
References
TASK-06-04, TASK-6-4, TASK-RR LSO5-81-12-063, LSO5-81-12-63, NUDOCS 8112230515
Download: ML17194A357 (18)

Text

{.

Docket No.

50-237 LS05*81-12-063 Mr. L. Del George Director of Nuclear Lic~nsing Connnonwealth Edison Company

• Post Office Box 767

~i Chicago, Illinois 60609

~:

Dear Mr. Del George~

1'.

o'ecember 18' 1 981

SUBJECT:

FORWARDING DRAFT EVALUATION REPORT OF SEP TOPIC VI-4, CONTAINMENT ISOLATION SYSTEM FOR THE DRESDEN NUCLEAR POWER PLANT, UNIT 2 I

1:

' t.

Enclosed is a copy of ocir draft evaluation of SEP Topic VI-4 1 Containment I~olat-ion System.

This :!assessment compares your facf11ty, as described in Docket No. 50-237, with the criteria currently used by the regulatory staff for lfcensing new!1faci1fties.

Please fnform us if your as-built facility differs from th,e licensing basis assumed in our assessment.

J Two of the more significant issues contained in the conclusion are use of manual and locked open valves as isolation valves; and use of check valves as isolation valv,,es outside containment. Both of these items (lppear not to comply wft:h the explicit wording of the regulat&on and

• no other acceptable defined basis could be determined from the informa-tion provided.

In addit'.fon,* sufficient information was not avaflabl e for us to evaluate if th'e automatic isolation valves indeed took the position of greater saf~tY upon loss of actuating power, as required by Appendix A to 10 CFR SO.Ii.

L<

I To enable us to perform 'our assessment of the deviations identified in this report, we w111 'need the defined basis upon which the specific isolation configurations!! at the Dresden Unit 2 Plant were judged to be eo<f ac~eptable @y you.

Plea,~e provide thfs information as a part of your J

.s connnents on this report.:

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, 8112230515 811218 PDR ADOCK 050001 237 P

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OFFICE................................................. *......................... ************************

SURNAME~... ;....................

DATE.........................

NRG FORM 316 (10-60) NRCM 0240

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... ~(*....................... *******************

OFFICIAL RECORD COPY USGPO: 1981-335-960

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'I Conunents are required* wi;thfn 30 days of receipt of this 1 etter so that they may be included 1n our final report. This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes nee~ed to reflect the as-built conditions at your facflfty. This assessme~t may be revfsed fn the future if your facility design is changed or ff ~RC crfterfa relating to thfs subject are modified before the integrated assessment fs completed.

Enclosure:

As stated cc w/ enc 1osure1

• _See next page Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensfng J c:.L ifj J. Lanle

\\~ IV.~

J. ~dRick OFFICE~ ******* ~~P..~

SURNAME~ ** ~.~.~~~.~.=. {!).~

DATE ***** 1.. ~!.5.7.~L....

. SEP:S

• • • sHeriiiaiin

!t<IRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1981-33:.-960

I.

. I.

INTRODUCTION CONTAINMENT SYSTEMS BRANCH EVALUATION REPORT ON SEP TOPIC VI~4 CONTAINMENT ISOLATION SYSTEM FOR THE DRESDEN NUCLEAR PLANT, UNIT 2 DOCKET NO. 50-237 Th;s report is a compilation of those aspects of the Containment Isolation Sys-tem for the Dresden Nuclear Power Plant. Unit 2 (Dresden 2) whfch do not meet the current licensing crf terfa fn use.by the NRC.

The f nfonnatfon for this report was obtained from a search of NRC docket ;nfonnatfon, including the

• ~

following princ;pal dbcuments:

A)

Dresden Nuclear Pbwer Station, Units 2 and 3, Safety Analys;s Report, Volume 1.

B)

Letter dated February 25, 1980 from D. L. Peoples, CE, to H.* R. Denton, NRC, "Comp 1 i ance with Category A Items of NUREG-0578 at Dresden 2/3 and Q~ad Cities Statibn.N C)

Attached pip;ng and instrumentation drawings included as part of Inser-v ice Inspection and Testing Programs for Dresden, Units 2 & 3, July 31,*

1978.

D)

Letter dated August 18, 1980 from R~ F. Janecek, CE to Gus Lainas, NRC, NSEP Topic XV-16. ".

E)

Appendix A to Operating Lfc:ense DPR-19, "Technical Specifications and Bases for Dresden Nuclear Power Station, Unit 2," May 5, 1971.

\\'

\\

\\ II.

REVIEW CRITERIA The* safety criteria used in the current evaluation of the containment isola-tion system for Dresden 2 are contained in the following references:.

1) 10 CFR Part SO, Appendix A, General Design ~riteria for Nuclear Power Plants (GDC 54, 55, 56 and 57).

2)

NUREG-75/087, Standard Review Plan for the Review of Safety Analy-sis Reports *for Nuclear Power Plants (SRP 6.2.4, Containment Iso-1 atf on System).

3)

Regulatory Guide 1.11, Instrument Lines Penetrating Primary Reactor Containment.:

4)

Regulatory Guide 1.141, Revision 1, Containment Isolation PrCNisions I

for Fluid Systems.

III.

RELATED SAFETY TOPICS In order to avoid duplication of effort the review areas identified below.are not covered in this report since they will be reviewed under other topics or ongoing generic reviews.

However, they are related and essential to the com-pletion of the. reevaluation of the containment isolation system for Dresden 2.

They are: *

1)

III-1, Classification of Structures, ComJ)onents and Systems (Seismic and Quality)

2)

III-4.C, Internally Generated.Missiles

3) *II I-5.A, Effects of Pipe Break on Structures, Systems and Components Inside Containment

,, 4) 111-5.B, Pipe Bre~k Outside Containment

5) 111~6, Seismic Design Considerations

6)

III-12~ Environmental Qua1fffcation of Safety Related Equipment

7)

VI-6, Containment Leak Testing 8).

VII.;.2, Engineered Safety Feature System Co~tro~ Logic and Design

9)

VIII-2, Onsite Emergency Power Systems - Diesel Generator

10)

VIII-4, Electric Penetrations of Reactor Containment

11)

NUREG-0737, Clar,ification of TMI Action Plan Requirements, Item II.L4.?, Containment Isolation Dependability

• 12)

NUREG-0660, NRC Actf on Plan Developed as a Result of the TMI-2 Accident,. Item II.E-.4.4, Containment Purging and Vent1ng Requirements

13)

NUREG-0803, Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping

- 4 IV Review Guideline~

The containment iso1atiO!? syste~ o~.a nuc1ear po"'er c1ant is ar. engineere~

safety feature that functions to a11olr0 thE norr.;a1 or er..ergenc.v passage o~

fluids through the containment boundary whi"le preserving the 1t'i1 ity of *the.

boundary to prevent or limit the escape of' fission products to the environ~

that may resu1t from postu1ated accidents.

Genera1 Design tritel'"ia 54, 55,

56. a*nd 57 of Appendix A to 10 CFR Put 50 pertain to the. containment 1so-1ation system of 1 nuc1ear power plant.

Genera1 Desi'gn Criterion* 54.establishes design and test r.equirements for the leak ~etection provisions, the fso1ation function and the containment capabiHty of the iso1ation barriers in lines penetrating the primary re-actor containment.

From the standpoint of containment iso1ation,* leak de-tection provisions should be capable of quickiy detectin9 and responding to a spectrum o.f postu1ated pipe break accident conditions..

To e*ccompHs.h this, diverse parameters shou1d be monitored to initiate the containment iso1ation function.

The parameters selected should assure a positive,

• rapid response to the developing accident condition. Th1s aspect of the containment isolation system review wi11 be addressed during the review of the post-TMI requirements approved for implementation, as stated in NUREG-0737 at Item II.E.4.2.

le-ak detection capability should also be provided at.the system level to alert the operator of t.he. need to isolate a system trai.n equipped with re-mote manua*1 isolation valves.

SRP 6.2..4, a-t Item II.6. b provides guidance in this regard.

5_

With respect to the design requirements for the isolation function, all non-essential systems shculd be ~utomatically isolated (with manua1 valves sealed closed), and valve closure times should be selected to assure rapid isolation of the containment in the event of an accident. The review of the classification of systems as essential or non-essential, and the auto-*

matic isolation provisions for non-essential systems by appropriate signals,

• will be addressed in conjunction with the review of the post-TMI require-ments as stated in NUREG-0737 at Item II.E.4.2. The closure time of the containment ventilation system isolatfon valves will be evaluated in.con-junction with the ongoi.ng generic review of purging practices at operatin~

I plants (see NUREG-0660 at Item II.E.4.4).

The electrical power supply, instrumentation and control systems should be designed to engineered safety feature criteria to assure accomplishment of the containment isolation function. This aspect of the review is covered under SEP Topics VII-2 and VIII~2: Also, resetting the isolation signal should not result in the automatic re-opening of containment isola~ion valves. *This will be addressed in conjunction with the review of the post-TMI requirements approved for implementation, as stated in NUREG-0737, at I tern I I. E

• 4

• 2.

With respect to the capabilities of containment isolation barriers in* lines penetrating primary co~tainmen't, the isolation barr*iers should ~ designed to engineered safety feature criteria, and protected against missiles. pipe

• whip and jet impingement. Typical isolation barriers 1*nclude valves, c*losed systems and blind flanges.

Furthennore, provisions should be made to pennit periodic leak testing of the isolation barriers.

e *. \\

The a~~~~acy of the.mis~ile, pipe whip arid jet i~pingement protection will be covered under S~P Topics 111-4.C, 111-5.A and 111-5.B.

The ac-ceptability of the design criteria originally used in the design of the contain~:nt isolation system components will be covered in SEP Topics 111-1, III-6 and 111-12.

The adeq~acy of the leak testing program will be covered under SEP.Topic

\\'l-6.

The acceptability of electrical penetrations will be covered in SEP Topic Vlll-4.

General Design Criteri; *55, 56 and 57 establis~. e).plicit *r~quirements

• for is~~2tion valving iri lines penetrating the co~tainment. Specifically, they eddress the nu~.~r Ind location of isolation valves (e.g ** reciun~ant i*

valvin; ~ith one located inside containment an~ the Qt~er located outside contain;:-.ent), valve actuation provisions (e.g., autorr.atic or rel"'IOtP manual.

• iso12tion va~ves), valve position {e.g., loc~ed c1osed, or the position of greater s~fety in the event of an accident or power failure), and va1ve tyoe (e.g., a simple check valve is not a penr:issable autorr.atic isolatfon va1ve out-.

side contain~ent).

1 GD: 55 and 56 a1so pennit containment isolation provision,s for lines pene-trating the primary containmen1 boundary that differ from the explicit re-quire:-.er.ts, provided the basis for acceptability is defined.

!~is provi.so is typically involed when establishing the contain~ent isolation require-ments for ~ssentia1 (f.e~. safety related) syste~s. or there is a clear im-prover:*.ent in safety.

.. * *Standard Review Plan 6.2.4.~ Itein II.6 presents guicielines __ f~r __ a:c~p~a~le alternate containment isolation provfsfons ~or certain class~s of lines.

Containment 1so1atfon provisions that are found acceptable on the *other defined basis" represent' confonnance with the *soc arid do not constftute exceptions.

e *.

\\ V.

EVALUATION

, The containment isol~tion provisions for the lines penetrating the.primary reactor containment of the Dresden Nuclear Power Plant, Unit 2 are tabulated in Table 1. This information was obtained from the documents referenced in**.

Section I~.. 'and questions sent to Cormnonwealth Edison on.April 23, 1981

{Re:

D.

• M. Crutchfield to J. S. Abel)~ There was insufficient informa-tion to complete Table l; therefore, the licensee should.provide the missing* information,.. and make any necessary correctfons.

The containment isolation 'provisions, as tabulated in Table 1, were evaluated against th.e requirements of General Design Criteria {GDC) 54, 55, 56 and 57 (Appendix A to l 0 CFR Part 50), and the supplementary guidance of Standard Review Plan (SRP) 6.2.4 (Co.ntainment Isolation System), where applicable.

Deviations from the explicit requirements of GDC 54, 55, 56 and 57, and the acceptance criteria of SRP 6.2.4 are summarized below:

1.

Insufficient administrative control;

2.

Insufficient leak detection capability on remote manual valves;

3.

Use of manual and locked open valves as isolation valves; and i

4.

Check valves as.isolation valves outside containment.

e. In addition, SRP Section 6.2.4.1.1.c states that the Containment Systems Branch review covers the verification of isolation valve positions in the normal, shutdown, post-LOCA and power failure modes.

The license.~ was requested to proyide this information* so that this review area could be completed.

To date, only the normal valve position has been provided.

The licensee is again requested to provide this information.

VI.

CONCLUSION The results of evaluation of the containment isolation system is best summarized by listing th~ areas of non-conformance to current licensing ctiteria, which wer~ provided abbve.

The remaind~r of this report

• gives a detailed discussion of these deficiencies. With these exceptions of the items identified above, the remain~er of the Containment Isolation System *for Dresden 2 listed in SAR Table 5.2.4 is found acceptable and in conformance with; current*: licensing standards.

ADMINISTRATIVE CONTROL All valves located between the inboard containment isolation valves or before the final outboard isolation valve (in the event there are none inside containment).should be locked closed to ensure the integrity of*

the piping between these valves. Typically, these valves are for test connections, vent lines, or capped branch lines. Listed below is a compilation of the valves in the Dresden 2 plant which need to be locked closed.

The available piping and instrumentation diagrams indicate that many of these valves are 11 normalry closed.

11 It does riot appear, however,

e.

that this designati~n meets the NRC defini~ion of "locked closed" beca~se there are no tags, mechanical lock devices or ~d~inistrat~ve procedures in exist~nce to prevent the valves from bei~g inadvertently opened.

Administrative control, as discussed fn SRP Section 6.2.4.II.~.f, need. to be,:-..~;f-~-~;~~~~-~-~:_?~;-~~e-~ -~-~--------~~=]

~-*----------*-'. --*--*--*****---- -* -- -*.-------*--1 VALVES REQUIRED TO BE LOCKED CLOSED PENETRATION X-105A,B,C,D X-106 X-107A,B X-108A X-1098 X-lllA,B X-113 X-115A X-116A,B.

X-122 X-130 X-144 X-145 x.:.147 X-149A,8 VALVE NUMBER Test Connections inside Con-tainment 220-5,6 220-103A,B; -104A,B; Valves on Hose Connection 1301-34, 35, sos, 506 1301-32, 33, 600, 601 1001-45A, 46A, 47A,B; -48A,8;

-90A,8,C; ~91A,8,C; -92A,8,C;

-206A,8,C 1299-7, 8; 1201-31, 32; Va1v~ on Capped Line 2301-16, 17 1501-23A,B; -24A,B 220-42, 43 1199-106, 107 301-96, 97 1501-298, 308 205-25, 26 1402-32A,B; -33A,B I

l.

i

X-150A X-3Q3A,8,C,D X-304 X-310A,8 X-3HA,8 X-312 1501-70A,8; -72A,8; -73A,8 2301-37, 3&; 2 Valves on Drain Line (Dwg.

11M-51")

1699-50, 51 1501-878; Valv~ on Capped End of Line 2-1522-14-LX 1501-40A,8; -41A,8 2301-418, 428 I

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--*-*-------*------r------**.

. ----~""'-"--

LEA.f< DETECT! ON SRP Section 6.2.4 *.II.<fib' & c states that remote manually operated valves located in Hnes in engfoeered s'afety features systems, engfoeered safety feature-related systems or used for safe: shutdown of the plant should be provided with leak detection c.apability fo~

1 leaks outside containment. Below, is a list of the remote manual val,ves at Dresden 2 which meet the description above and are required to have leak detection capability.

Upon q.uestioning the licensee, I

it appears that these va*lves do not have the requisite leak detection capability:

SYSTEM.

LPCI LPCl Core Spray Core Spray Reactor Bldg*. Closed Cooling Water System Ditto i

.PENETRATION VALVE NUMBER X-303A,B,C,D 1501-SA,B,C,D X-116A,B l501-22A,B X-303A,B,C,D 1402-3A,B X-149A,B 1402-25A,B X-123 3702 X124 3703

,.~ *--*-***~-- ******---* -------~*----*. __., --*-*

Manual Isolation Valves General Design Criteria (GDC) 55, 56 and 57, Appendix A of 10 CFR Part 50 state that unless it can b~ demonstrated acceptable on some other defined basis, isolation valves should be either automatic or locked closed.

A non-lock~d closed manual valve serving the purpose of a containment isolation barrier does not meet current regulations.

SYSTEM PENETRATION VALVE NUMBER Service Water Supply X-119 4337-500, 502; 1916-500 Service Air Supply X-120 4609-501, 11Mis-cellaneous 11 Valve (SR Table 5.2.4)

HPCI Condensate Drain X-312 2301-81 (Locked Open)

HPCI Turbi.ne.Exhaust X-317A 2301-74 (Locked Open)

.Furthermore, the last two manual valves, valves 2301-71 and 74, are locked open.

Therefore, they were. not considered isolation barrters.

CHECK VALVES AS ISOLATION VALVES GDC 55 and 56 state that a 'check valve alone outside containment cannot be con-sidered an isolation barrier.. The*feedwater penetrations, X-107A an.dB, however, do contain only single check valves outside containment.

Consequently, this does not meet current criteria.

In light of the safety significance of the feedwater penetrations the stiff beli~ves that an acceptable isolation b~rrier for this p~netration would consist of the check valve outside containment along with a remote manual valve.

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