NUCLEAR REGULATORY COMMISSION - Turner Broadcasting …i2.cdn.turner.com/cnn/2012/images/02/16/nrc.gl.89.16.pdf · v BOSTON EOISON CWPANY August 18, 1988 U.S. Nuclear Regulatoty Comnl

UNITED STATES NUCLEAR REGULATORY COMMISSION

WASHINGTON, D. C. 20555

September 1, 1989

TO: ALL HOLDERS OF OPERATING LICENSES FOR NUCLEAR POWER REACTORS WITH MARK I CONTAINMENTS

SUBJECT: INSTALLATION OF A HARDENED WETUELL VENT (GENERIC LETTER 89-16)

As a part of a comprehensive plan for closing severe accident issues, the staff undertook a program to determine if any actions should be taken, on a generic basis, to reduce the vulnerability of BWR Mark I containments to severe accident challenges. At the conclusion of the Mark I Containment Performance Improvement Program, the staff identified a number of plant modifications that substantially enhance the plants' capabi lity to both prevent and mitigate the conse uences of severe accidents. The improvements that were recommended include ? 1) improved hardened wetwell vent capabi 1 i ty , 2 improved reactor pressure vessel depressurization system re1 iabf If ty , 3 an alternative water supply to the reactor vessel and drywell sprays, and 4 updated emergency procedures and training. The staff as part of that I i effort also evaluated various mechanisms for implementing of these plant improvements so that the licensee and the staff efforts would result in a coordinated coherent approach to resolution of severe accident issues in accordance with the Commission's severe accident pol icy.

After considering the proposed Mark 1 Containment Performance Program (described in SECY 89-017, January 1989), the Commission directed the staff to pursue Mark I enhancements on a plant-specific basis in order to account for possible unique design differences that my bear on the necessity and nature of specific safety improvements. Accordingly, the Commission concluded that the recommended safety improvements, with one exception, that is, - hardened wetwell vent capability, should be evaluated by If censees as part of the Individual Plant Examination (IPE) Program. U i th regard to the recommended plant improvement dealing with hardened vent capabi 1 i ty, the Commission, in recognition of the circumstances and benefits- associated with this modification, has directed a different approach. Specifically, the Commission has directed the staff to approve installation of a hardened vent under the provisions of 10 CFR 50.59 for licensees, who on their own Ynitiative, elect to incorporate this plant improvement. The staff previously inspected the design of such a system that was installed by Boston Edison Company at the Pilgrim Nuclear Power Station. ' The staff found the installed system and the associated Boston Edisdn Company's analysis acceptable.

A copy of Boston Edison Company's description of the vent modification Is enclosed for your information. For the remaining plants, the staff has been directed to initiate plant-specific backfit analyses for each of the,Mark I plants to evaluate the efficacy of requiring the installation of hardened wetwell vents. Uhere the backfit analysis supports imposition of that requirement, the staff is directed to issue orders for modifications to install a reliable hardened vent.

Generic Le t te r 89-16 September 1, 1989

The staff believes tha t the avai lable information provides strong incent ive f o r i ns ta l l a t i on o f a hardened vent. F i rs t , i t i s recognized t ha t a l l affected plants have i n place emergency procedures d i rec t ing the operator t o vent under ce r ta in circumstances (pr imar i ly t o avoid exceeding the primary containment pressure l i m i t ) from the wetwell airspace. Thus, incorporation o f a designated capabi 1 i t y consistent w i th the object ives o f the emergency procedure guidelines i s seen as a log ica l and prudent p lant improvement. Continued re l iance on pre-exist ing capabi l i ty (non-pressure-bearing vent path) which may jeopardize access t o v i t a l p lant areas o r other equipment i s an unnecessary complication t ha t threatens accident management strategies. Second, implementation of r e l i a b l e venting capabi l i ty and procedures can reduce the 1 i kel i hood of core me1 t from accident sequences involv ing loss o f long-term decay heat removal by about a factor o f 10. Rel iable venting capabi l i ty i s a lso beneficial, depending on p lant design and capabi l i t ies, i n reducing the l i ke l ihood o f core melt from other accident i n i t i a t o r s , f o r example, s ta t ion blackout and ant ic ipated transients wlthout scram. As a mi t igat ion measure, a r e l i a b l e wetwell vent provides assurance o f pressure r e l i e f through a path w i th s ign i f icant scrubbing o f f i s s i o n products and can resu l t i n lower releases even f o r containment f a i l u r e modes not associated w i th pressurizat ion (fee., l i n e r meltthrough). Final ly , a r e l i a b l e hardened wetwell vent allows f o r consideration o f coordinated accident management strategies by providing design capabi l i ty consistent w i th safety objectives. For the aforementioned reasons, the s t a f f concludes tha t a p lant modif icat ion i s h ighly desirable and a prudent engineering so lu t ion o f issues sur,rounding complex and uncertain phenomena. Therefore, the s t a f f strongly encourages 1 i censees t o imp1 ement requ is i te design changes, u t i 1 i r i n g port ions o f ex is t ing systems t o the greatest extent pract ical , under the provisions of 10 CFR 50.59.

A; noted previously, f o r f a c i l i t i e s not e lect ing t o vo lun ta r i l y incorporate design changes, the Commission has di rected the s t a f f t o perform plant-specif i c b a c k f i t analyses. I n an e f f o r t t o most accurately r e f l e c t p lant spec i f ic i ty , the s t a f f herein requests tha t each licensee provide cost estimates f o r implementation o f a hardened vent by pipe replacement, as described i n 'SECY 89-017. I n addition, licensees are requested t o indicate the incremental cost o f i n s t a l l i n g an ac independent deslgn i n comparison t o a design re ly ing on a v a i l a b i l i t y o f ac power. I n the absence o f such Information, the s t a f f w i l l use an estimate o f $750,000. This estimate i s based on modif icat ion o f preval ent ex is t ing designs t o bypass the standby gas treatment system ducting and includes piping, e l ec t r i ca l design changes, and modifications t o procedures and training.

The NRC s t a f f requests tha t each licensee w i th a Mark I plan t provide n o t i f i c a t i o n o f i t s plans f o r addressing resolut ion o f t h i s issue. If the 1 icensee e lec ts t o voluntarily proceed w i th p lant modifications, i t should be so noted, along w i t h an estimated schedule, and no fur ther information i s necessary. Otherwise, the NRC s t a f f requests t ha t the above cost information be provided. I n e i t he r event, i t requests t ha t each licensee respond w i t h i n 45 days o f rece ip t o f t h i s l e t t e r .

Generic Let ter 89- 16 -3 - September 1, 1989

This request i s covered by Of f i ce o f Management and Budget Clearance Number 3150-0011, which expires December 31, 1989. The estimated average burden hours are 100 person hours per licensee response, includfng searching data sources, gathering and analyzing the data, and preparing the required le t ters . These estimated average burden hours per ta ln only t o the i d e n t i f i e d response-related matters and do not include the tfme fo r actual Implementation o f the requested actions. Send comments regarding t h i s burden estimate o r any other aspect o f t h i s co l lec t ion o f information, includfng suggestions f o r reducing t h i s burden, t o the Record and Reports Management Branch, D iv is ion o f Information Support Services , O f f i ce o f Information Resources Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555; and t o the Paperwork Reduction Project (3150-0011), Of f i ce o f Management and Budget, Nashington, D.C. 20503.

If you have any questions regarding t h i s matter, please contact the NRC Lead Project Manager, Mohan Thadani , a t (301) 492-1427.

S t ncere ly .,

~ s s b c f ate Of f ice o f

Di rector f o r Projects Nuclear Reactor Regulation

Enclosures: 1. Description o f Vent

Modif icat ion a t the P i lg r im Nuclear Power Stat ion

2. L i s t o f Most Recently Issued Generic Let ters

w Enclosure 1

BQnrWVlgWUMT Pdgr~m Nuclear Power Stat~on

Rocky Hdl Road

. . Plymouth. Massachusetts 02360

I ' 8 ,

Ralph G. Bird Sentor v~ce Presiaert - Nuclear

BECO 88- 126 A u g u s t 18, 1988

U. S. Nuclear Regu'latory Comml sslon Document Control Desk Washington, DC 20555

L l cense DPR-35 Docket 50-293

REVISED INFORMAT ION, REGARDING PI LGRIH STAT ION SAFFlY ENHANCEMFNT PROGRAM

Dear S l r :

Enclosed i s a descrlpt lon o f a revlsed deslgn for . the Dl rect Torus Vent System (DTVS) that was descrtbed I n the *Report on P11Qrim Statlon Safety Enhancements" dated July 1, 1987 and transmltted t o the NRC wlth Hr. Blrd's l e t t e r (BECo 87-111) t o Hr. Varga dated Ju ly 8, 1987. Thls rev ls lon supersedes I n I t s en t i re ty the Section 3.2 Included I n the Ju ly 1, 1987 report.

On March 7, 1988 Boston Edtson Company (BECo) personnel met w l th Dr. Murley, M r . Russel 1, and Dr . Thadanl and provlded a tour o f SEP md l f l ca t l ons and an Informal ptesentatlon o f the quant i f lcat lon o f competing r l s ks assoclated w l t h venting the contalnment and conclusions drawn from these results. Thls ptesentatlon provlded BECo the opportunlty t o respond t o questions posed under Item 1 Section 3.2 - "1nstal.latlon o f A Dlrect Torus Vent System 7DNS)" I n Mr. Varga's l e t t e r t o Mr. B l rd o f August 21, 1987 * I n l t l a l Assessment o f P l lgr lm Safety Enhancement Pr~grarn~~. The materlal presented was made aval lable t o the resldent Inspector and was Included as Attachment 11 I n NRC 1nspectlon.Report #88-12, dated May 31, 1988.

As you are aware from plant lnspectlons we have Ins ta l led the DTVS plplng and portlons o f re lated control wlring. 'Currently, the DTVS 1s Isolated from the Standby Gas Treatment System (SBGTSI by b l i nd flanges i ns ta l l ed I n place o f Valve 40-5025 and the DTVS rupture dlsk. This configuration was Inspected by NRR I n the performance o f a technlcal revlew whlch focused on System, Mechanlcal Deslgn and Structural Deslgn Pssues. The revlew took place on Hatch 2-3, 1988 as documented I n NRC Inspection Report rOr88-07, dated Hay 6, 1988 and detemlned the I ns ta l l a t l on conflguratlon t o be acceptable. We now plan t o remove these b l l nd flanges and proceed wi th l ns ta l l a t l on o f Valve A0-5025 and the DNS rupture dlsk. He conclude the valve and rupture d lsk provlde equlvalent physlcal l so la t lon o f the D N S pip ing from the SBGTS and appropriately ensure the operatlonal Integrity o f the SBGTS under deslgn bssls accldent condltlons. Followlng completion o f t h i s work, we w l l l perform a local leak ra te t es t t o v e r l f y that Valve AO-5025 I s acceptably leak t l g h t uslng the same method previously u t l l l z e d i n tes t lng the b l l nd flange. He also plan t o complete a l l remalnlng e lec t r i ca l work on the OTVS i n accordance wlth the revlsed deslgn.

v BOSTON EOISON CWPANY August 18, 1988 U.S. Nuclear Regulatoty Comnl sslon

Page 2

On the basis o f the revised Sectlon 3.2, we conclude that the OTVS design as described I n the enclosure does not requlre any change t o tbs Techntcal Speclf lcatlons and tha t we can proceed w l t h I ns ta l l a t i on w l thout p r l o r NRC approval . Please feel f ree t o contact me o r Mr . 3 . E. Howard, o f my s t a f f a t (617) 849-8900 i f you have any questions pertaining t o the design deta i l s o f the OTVS.

Attachment: Sectlon 3.2 Revlslon 1 * ~ n s t a l l a t i d n 01 A Direct Torus Vent System (DTVS)".

cc: Mr . 0. McDonald, Project Manager Dl v l sl'on' o f Reactor Pro3 ects 1/11 Of f l ce o f Nuclear Reactor Regulatlon U. S. Nuclear Regulatory Comi ssion Mail Stat lop PI-137 Hashlngton, D.C. 20555

U. 5. Nuclear ~ e g u l a t o r y Comlsslon Region I 475 A1 1 endale Road King o f Prussia, PA 19406

Senlor NRC Resl dent Inspector Pi 1 grlm Nuclear Power Station

Attachment to BECo Letter 88-126

Section 3.2 Rtvlrlon 1 @Inrtallatlon O f A direct Torus Vent System ( D M ) @

pages 14, IS, 16, 17, 18, 19, 19A, 198

Thl s deslgn change provides the a b l l l t y for direct : venting o f the torus' to the main f tack. . Cbntalnment

venting i s one core damage prevention strategy u t i l l zed i n the SW!? .Ownets Group Emergency Procedure Gui del tnes (EPGs) as previously approved by the NRC and I s requtred I n p l ant-spec1 f l c Emergency Operating Procedures (EOPs) . The torus vent l i ne connecting the torus t o the maln stack w i l l provide an alternate vent path for Imp1 ementlng EOP tequi rements and represents a slgnlf lcant improvement re lat ive t o exist ing plant vent capablli ty. For 56 psi saturated steam condi tlons i n the torus, apporoximately 1% decay heat can be vented.

Thls deslgn change (Figure 3.2-1) provides a dl rect vent path from the torus t o the maln stack bypassing the Standby Gas Treatment System (SBGTS). The bypass 1s an 8* 1 ine whose upstream end I s connected t o the pipe

. between primary containment isolat ion valves A015042 A b 8. The downstream end o f the bypass I s connected t o the

'

20" maln stack 1 ine downstream of SBGTS valves AON-108 and AON-112. An 8" butter f ly valve (AOI5Q25). which can be remotely operated from the maln control room, i s added downstream o f 8" valve AO-5092B. Thts valve acts as the primary containment outboard 1 solatlon valve for the d l rect torus vent 1 ine and w l l l conform t o NRC requirements for sealed closed isolat lon valves as defined I n NUREG 0800 SRP 6.2.4. The new pipe i s ASME I11 Class 2 up t o and Inclusive of valve AO-5025. Test connec tlgns are provided upstream and downstream o f AO-5025.

The design change replaces the existing AC solenoid valve fo r AO-5042B w l t h a DC sol en01 d valve (powered from essential 125 v o l t DCh t o ensure operabill t y w i thout dependence on AC power. The new Isolat ion valve, A0-5025, I s a1 so provided w l t h a DC solenoid powered from the redundant 125 vo l t DC source. Both o f these valves are normally closed and f a l l closed on loss o f electr ical and pneumatic power. One Inch n l trogen 1 lnes are added to provide nitrogen t o valves AO-5042B and A0-5625. New valve A04025 w l 11 be control 1 ed by a remote manual key-locked control swl tch. During normal operation, power t o the A0-5025 DC solenoid w i l l also be d l tabled by removal o f fuses I n the wlrlng t o the solenold valve. Thls sat is f ies NUREG 0800 SRP 6.2.4, Containment Is01 at ion System acceptance c r i t e r l a fo r a sealed closed bar t i er. An addl t lonal fuse w l l l be ins ta l led and remain I n place t o power valve status indication for A0-5025 i n the main control room.

-14- Rev. 1 (7/25/88)

I L' '. -. .d

New 8" vent pipe (8"-HBB-441, Includlng valve A 0 1 5 0 2 5 1s safety related. Vent plplng downstream o f A01W25, Including SBGTS discharge plplng t o main stack, I s also safety related. A l l safety related plplng w i l l be supported as Class I. N l trogen piping i s non-safety related and w i 11 be supported as Class II/I.

The interpretation of the Class I l l 1 desfgnatlon through t h i s report 1s given below:

A1 1 Class I1 I t e m f which have the potential t o degrade the In tegr i ty of a Class I 1 tern a t e analyzed. Such Class I1 Items do not requlre dependable mechanical or e lectr ical funct lonal l ty during SSE, only that a l l o f the f o l lowing condl tlons prevail :

1. The Class I1 Items create no missiles whlch impact unprotected Class I Items safety functions.

2. The Class I1 I tern does not deform I n a way which would degrade a Class I I tern.

3. If the Class I1 t t e m fal ls, then the Class I I t e m 1s protected against the f u l l Impact o f a l l missiles ,

generated by the assumed fa i lu re o f Class I1 items.

A1 1 e lectr ical portions of th ls deslgn are safety related except f o r the indicating l ights on the HIHIC panel C904, the tI e-ins t o the annunciator, and Interface w l t h the plant computer.

The torus purge exhaust 1 lne Inboard I solatlon valve A0150420 and the assacfated BY plpe are the components of the CACS affected by the deslgn mod1 flcatlon. HI t h Incorporation o f the subject modltlcatlon, the CACS w l l l depend on both essenttal AC ( for valve A015042A) and essential QC (for AO-50428) t o perform I t s putgi ng function.

The new 8" torus vent l l n e w i l l be connected to exlst lng 8" CAG plplng between valves AO-5042B and AO-5042A.

Rev. 1 (7125188)

The SBGTS fan out le t valves (AOQI-108 and AON-1121, ductwork from these valves t o the 20" l i ne leading t o the main stack, and the 20" l i n e leading t o the main stack are the components o f th ls system affected by the proposed change.

Valve AON-108 I s normally closed, fall-open. Valve AON-112 i s normally closed, fall-closed, and these valves are provided wlth essential DC power and local safety related a i r supplies.

prirnarv Containment Isolat ion System (m Valve A0-50428 I s affected by the change from AC t o DC power fo r the solenold and by replacement o f the exlstlng a l r supply w l t h

. n i trogen . The addl t i on o f contal nment outboard Isolat ion valve W-5025) w i l l not affect the PCIS.

Prfmarv ~ m i ! . m e n t e m ~~~ Valve A0-5025 acts as the p r i m r y contal nment ' outboard isolat ion valve fo r the di rect torus vetit 1 ine and wf 11 conform t o HRC requl rements f o r sealed closed l solatlon valves as deft ned

. I n NUREG 0800 SRP 6 A 4 .

3.2.3.2 v Functlans of Affected Svstems/Cmanents

ThEs system has the safety functlon o f rhduclng the poss lb l l l t y o f an energy release w i t h in the prlmary containment from a Hydrogen-Oxygen reactton f o l loving a postulated LOCA combl ned w l t h degraded Care Standby Cool t ng System.

This system f i l t e r s exhaust a l r from the reactor bu l l d l ng and d l scharges the processed a i r t o the maln stack. The system f i l t e r s particulates and lodlnes from the exhaust stream i n order t o reduce the level o f airborne contaml nation re1 eased t o the envi tons v ia the maln stack, The SBGTS can also Pi1 t e r exhaust a l r from the drywell and the suppresslon pool.

-17- L Rev. 1 (7/25/88)

P n c n t Iso lat ion S v w

This system provi des tlmely protection against the onset and consequences o f deslgn basis accidents involving the gross release o f radioactive materials from the primary contalnment by In1 t i a t i ng automatic Iso lat ion of appropriate pipe1 ines which penetrate the prlmary containment whenever monitored varlabl es exceed pre-sel ected operational l imi ts .

Prlmarv Cantal nment Svstem

The primary containment system, I n conjunction with other safeguard features, 1 i m i t s the release of f ission products i n the event of a postulated design basis accident so that o f f s t te doses do not exceed the guideline values o f 10 CFR 100.

3.2.3.3 potent ial Effects on Safety funrt iont

h e r l t Control Svs-, SthllSlbY, s T r e a t m e n t . P r m r v C o n t a l e l a t i on Svstem and Primary Containm-

The improvements change the -50428 sol enoi d control from AC to DC enabl lng i t t o open (from i t s normally closed positfon) with no dependence on AC power ava t lab i l l ty. The exist ing a i r supply t o A0-50428 1s being replaced by n l trogen.

Ductwork a t the out1 e t o f the SETS i s replaced w i t h plpe and the new vent l i n e i s connected t o the 20" l l n e a t the out le t o f the SBGTS.

Addltion o f a new 8" vent l l n e with contalnment Iso lat ion valve AO-5025 off. the ex1 st ing torus vent 1 ine could introduce a flow path under deslgn basis condlttons that could vent the contalnment d l rec t l y t o the stack bypassing the SBGTS .

3.2.3.4 Fnalrsls o f Effects on S-tv Fun-

An analysis o f the effects on the safety functions o f CACS, SBGTS, KIS and PCS for the ins ta l la t ion o f the d i rec t torus vent i s described as follows:

The change from AC t o OC control and the rep1 acemen t s o f a1 r w i t h n i trogen on A0150428 does not adversely af fect the a b i l i t y to open A0-504233 when the contalnment I s be1 ng purged, o r t o iso late under accident condl tlons.

-1 8- Rev. 1 (7 /25188)

u -

L/ t The modlflcatlons to the ductwork and 20' line

leading to the mln stack do not affect the . . desl gn basls safety function of any of the

safety related systems.

During normal plant operatlons, the CACS and t h e SBGTS do not use the torus 20' purge and vent llne to perform t h e l r safety functions. The contafnment 1 solatlon valves are In the1 r normally closed posl tlon, thus aralntal ning prlmary contalnment boundary I ntegrl ty.

There are no adverse affects on the primary contalnment system by the addttlon of the DTVS. Valve A015025 wlll conform to NRC

. crlterla for sealed closed isolation valves as deff ned in NUREG 0800 SRP 6.2.4 and will not affect deslgn bads accldents. Use of the D T S wlll be l n accordance wf t h the contal nment venting provl sionr of EPGs as approved by the NRC and controlled by EOPs In t h e sane manner as other exlstlng contalnment vent paths. Me

. effects on the torus of the new 8' plplng and A0-5025 have been evaluated for Mark I program loadings, uslng ASHE BPVC Section I I I crl teria. The remalnlng. pi ping lncludin t h e - rupture dlsk was evaluated usIng ANSI 03 .1

' requf rements. 0

During pl.ant startup and shutdown (non-emergency condi tlon) when the purge and ven t 1 lne l s In use, valve AO-5025 remains closed. In addltton, the rupture disk downstream ~f valve A015025 w i 11 provlde a second post t ive means of preventl ng 1 eakage and prevent dlrect release up to the stack during contalnment purge and vent a t plant startup or shutdown.

Durlng contalnment htgh pre$sure condltlons, t h e torus main exhaust llne i s automatically Isolated by the PCIS. There f s no chan e to the ex1 s t i ng prlmary containment Isolat 1 on system funcf ton for A01504U or A015042B. The sealed closed posl t l on of valve A015025 and the add! tlonal assurance added by the rupture dlsk downstream w l 11 prevent any Inadvertent dl scharge up the" stack for a1 1 derlgn bash accldent condi tlons ..

3.2.3.5 Chanae Ev-rv Con- .

Installatlon of the D N S does not adversely affect the safety functions of the CACS, SBCTS, PCIS or the Integrt ty of prloary contalnment or any other safety related systems.

-1 9- Rev. 1 (7125188)

Use of the D N S w i l l be I n accordance with the contalnlnent ventln provisions o t Effis as approved by the NR ! . and controlled by EOPS i n the same manne? as o t h w m l t t i n g containment vent paths. The D N S provides an Improved contal nment ventl ng capab1l.l t y for decay heat removal uhlch reduces potentla1 onslte and o t f s l t e Impacts re la t ive to the exlstlng contaf nmant ventlng capablll ty.

Rev. 1 (7/25188)

DIRECT TORUS VE-M