ML20196F327

Forwards Util Biweekly Status Rept for Period 870604-17

ML20196F327
Person / Time
Site:	Pilgrim
Issue date:	06/19/1987
From:	Wiggins J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	Collins S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20150E198	List: ML20150E195 ML20195G141 ML20196F053 ML20196F242 ML20196F264 ML20196F327 ML20196F336 ML20196F342 ML20196F348 ML20196F412 ML20196F418 ML20196F424 ML20196F430 ML20196F443 ML20196F452 ML20196F459 ML20196F469 ML20196F474
References
FOIA-88-198 NUDOCS 8812120262
Download: ML20196F327 (11)

Text

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/* " %,, UNITED STATES 1 NUCLEAR REGULATORY COMMISSION

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5'  ! OFFICE OF GOVERNMENTAL AND PUBLIC AFFAIRS, REGION I

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,/ 631 Park Avenue, King of Prussia, Pa.19406 Tel. 215 337 5330 No. I-87-91

Contact:

Karl Abraham June 23, 1987 NOTE TO EDITORS AND STATION ASSIGNMENT EDITORS James T. Wiggins, Chief of the Reactor Projects Section that inspects Pilgrim, has issued a status report summarizing activities of the inspection staff during the period June 4 - June 17, 1987.

The report is attached.

June 19, 1987

'Jocket No. 50-293 MEMORANDUM FOR: Samuel J. Collins, Deputy Director Division of Reactor Projects FROM: James T. Wiggins, Chief Reactor Projects Section 1B

SUBJECT:

PILGRIM STATUS REFORT FOR THE PERIOD JUNE 4. - JUNE 17,1987 Ent% sed is the Pilgrim bi-weekly status report from the NRC Resident Of fice at Pilgrim. Three NRC resident inspectors and two region-based inspectors monitored activities at the plant during the report period.

These status reports are intended to provide NRC management and the public with an overview of plant activitie', and NRC inspection activities. Subsequent inspection reports will address many of these topics in more detail.

(Original Signed By)

James T. Wiggins, Chief Reactor Prsjects Section IB (more) 8812120262 881208

$NS -198 PDR

9 l 2 ENCLOSURE PILGRIM STATUS REPORT FOR THE PERIOD JUNE 4 - 17, 1987

1. 0 Plant Status i As of 8:00 a.m. on June 17, 1987, spent fuel pool temperature was about j 60 degrees Fahrenheit. All fuel has been removed from the reactor vessel

and transferred to the spent fuel pool.

i 2.0 Facility Operations Summary 1

1 The plant has been shutdown for maintenance and to make program 4

improvements since April 12, 1986. The reactor core was completely de-t fueled on February 13, 1987. The licensee is presently performing exten-

! sive prever.tive maintenance and modifications of plant equipment. Upon 4

completion of certain modifications, the licensee plans to begin fuel

) reload.

l 3.0 Emeroency Notification System (ENS) Reports

! During this period, the licensec made two reports to the NRC pursuant to l 10 CFR 50.72.

i Unexpected Secondary Containment Isolation On June 7,1987, an unexpected secondary containment isolation signa! was i generated during daily surveillance testing of the refueling area exhaust I vent radiation monitors. While testing the upscale trip on the ' A' chan-i nel, the 'B' channel spuriously tripped causing the isolation signal. All systems responded as designed. The Standby Gas Treatment System fans were tagged out for maintenance and thus did not start. The licensee is inves-tigating the source of the spurious signal.

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Inoperable Fire Hydrant l On June 11, 1987, a fire hydrant and hose station required by the station j Technical Specifications was declared administrative 1y inoperable during a performance of the routine monthly visual surveillance check. The hydrant is located outside the process building but inside the protected area.

The hydrant was declared inoperable when it was discovered that less than the required length of fire hose was installed. Ths resident inspectors are following up on the licensee's causal analysis anc corrective actions.

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1 4.0 Items of Spec.ial Interest Containment Spray Header Internal Corrosion and Spray Nozzle Blockage i i I l On June 10, 1987, during implementation of a planned system modification,  ;

the licensee discovered significant amounts of rust in the drywell primary  !

. containment spray ring headers and nozzles. Primary containment spray in t the drywell consists of two independent eight inch ring headers located at ,

different elevations that can be used to spray water to cool the drywell. [

! The drywell spray system is not required to be used in response to large i break accidents, cut may be useful in dealing with small break LOCA scen-i arios. A similar containment spray system is used it, conjunction with [

] another cooling system to cool the torus. Reeoval of 104 nozzles from the  !

) upper drywell header revealed a large amount of rust flakes in the noz-  !

1 zies, and similar material distributs.d throughout the header approximately '

l l one-half inch deep. Sixty-four nozzles have been removed from the lower i j header with similar rust material identified in the header. Smaller quan- {

j tities of rust were found in the lower header spray nozzles. The licensee (

! is evaluating the nozzles to determine whether the presence of the rust [

I would have significantly affected the ficw rate and ' effectiveness of the l sprays.

I 5.0 NRC Staff Status During the Period l l Three NRC Resident Inspectors monitored plant activities between June 4, l and June 17, 1987. The inspection staff at Pilgrim during the report

] period consisted of the following: i l

Martin McBride, Ph.D. --- Senior Pesident Inspector j a

i Jeffrey Lyash --- Resident Inspector f j Tae Kim --- Resident Inspector  ;

l An NRC Region I specialist inspector was on site during this period to inspect non-licensed personnel training programs. Also, a regional

! project engineer assisted the resident inspectors during the week of j June 1, 1987.

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JUL 06 '87 13:46 WRC KING OF PRUSSIA P02 MAJOR FLANT CESIGN L64ANGE5 Al PILORIM DURING RFO=e7 FLANT CESIGN CHANUE No. CESCRIPTION 64-7n ANALOG TRIP SYS1EM - Risk of Inadvertent testing related scrams ned led to identt f tcation of 73 mechani cally actuated pressure, level, and differential pressure switches to be replaced with analc3 tr ansmi tt ers , master and slave trip untts and output relays. 44 of these swi t 6tews are associated with RFS/PCIS, 26 with ECCS, and 3 are none-class 1E. (Pre-op testing ongoing) 05-6/ REACTOR WATER LEVEL INGTRUMENTAT!DN MODIFICATION -

hoplacement of the esteting Yarway beated reference columms an the drywell with new cold reference twgs outside the drywell to reduce level indication wtrueo caused Qy high drywell temperatures. (NRC Generic Letter 84-23)

(>aetd nsplementat1on tw.ap!wLei d' s AEACTOR SUILDING $(CONLVnMY LONTAINNENT IF.0L ATION DAMPERS AEFLACEMENT -

Repiscemt t of the outstang commerctal we ade dampers in ventillation air intake and exhaust ducts with new nuclear grade campers. The new dampars have metal

!!nkage Anstead of nylon gears to ellminate tne wearing problems and failures. (Fre-op testing ongoing)

G5-57 ELETROLYTIC MYDRDGEN WATLH LHEMIETRY SYSTEM ( HWC S )

This modification includes installing a permanent hydrogen injectton system to inject hydrogen to the reactor coolant,. win the feedwater system, to reduce tne dissolved ow ygen concentret t on. The hydrogen injection system utillses electrolytical generetton of hydrogen and osygen emette. (Fleld i mpl ement a ti on ongoing l

67. ~ 0 9 H8C1 EanAunt vwyuuM FhLAVER . This modt f ttation adds A vecuum rettef line to the HPCI turbane OMhaust pnpe te ,

reduce the potential for covere hydrodynant e transs en's in the HPCI turbtne e=haust line. The new lane will i n c !'u d e t=u vetuum relief cneck valves, two normal l y open

, MOVstFCISt, anc two normally epen manual isolaton valves.

(Aeft NAC sgri to.;gsfe:.;o) - Field supletentatior ongoing

t - ? .' swm CCNTAINMfNT ISVLartyN vnLvES REFL ACEMENT -

ina avnstang gat, and globo val <es(MOV 1001 ' ;9 A ,000. h+ . .'- W installed in RHk trains A t. b are being replaces by gate valw$s witn doubtw dite paralled seat and g!cce valves wn s c r.

4e manufactured tw mnem S t e n ng en t r.t a t er n a l and inspec t n on requnee*,ents the Lbe entstang valvet. Yne existing vanvet nas e,swant.e seat leelece, eegeatet i kl +ailure, er t:s t e wear en velve seatteiv wurfers, packt.g leanage, an0 becy to beenet e...ew ...! r a n<3 lea 6 age.

<rgese imple,,entation omgetor) 1 I;t

. .IUL 06 W 13146 HRC KING OF PRUSSIA P03 FLANT SION '

ute 4L 60. '

DE SC A 4 P i 4 6.N 06-15 HCU SCRAM DibLHAbOE RISER SUFFChi MQDIFICAT!QN -

This design ahange removes the existing, celgtnal 3.way(p,y,

1) pape supports for the 3/4" scram disenarge raserett45) and replaces each suppor t wa tn . wa y ( x , Z) supports to recuce thermal stress on the scram discharge riser lines.

tFleid implementatton ongotng) 86-51 DIRECT TORUS VENT (bEF) -

This modif ication adds an 9" torus vent line connecting the t orus to the main stac k wttch will provide a flow path for r et t eva ng primary containmor t pressure to presrht 9verpressurtsatton anc subsequent f ailure of the primary containmont during a severe acen dent.

(> told implementation ongoing) 86."2A hkPLACEMENT OF r.ONT4!NMENT CrnAY LEADER CAPS /NO22LES (EEE) -

The modificativo &nstudes replacement of 104 upper anc 144 lower containment spray header nearles. The replacement nogales are ident ical to the existing nogales except that the replacement fogjet nogales have one open spray cap anc sta blanked off spray caps, whereas the entsting fogjet nnPPlpm have all eawsm Papg upon. Shw 9WdWees Cpray (IGH w&11 decrease the possibility of damagtng the containeont structure by sudden decompression.

(Field implementation ongoing)06-52P FIAT WATER TIE-IN TO RHR (SEP) -

Installation of a piping crosstle between the fire water system and the RHR system to provide redundant sources of water for upper RNM containment mue my hvadur, torus spray header, amator Nyv injection. (fl el d implementation ongoing) 96.SJC IHCTnLLnTION Dr (OCPI D I CDCL F I M E P ul le - R wp t ab w mwn s 09 Ene entstang abandoned construction fire water pu mp located wast uf the saattnyaro witn a mar eteset tare pump. The SEP diesel fire pump will be used to puep city water supply to the fire water system. Thr ough the crosette between the firm water system and the RHR system, the SEP diesel fire

' pump and the eststing fire pu*ps wit! De able to provide enough discharge. pressure to pump fire wa*.2- so the upper RHR contatnment spray header , torus spray header, and/or RPV injection af needed. (Fland implementation scheduled ,

to began on July 3, 1987)66-52D FL4v O!L 1RRNSFER SYSTEM FCA (SEF) DIESEL FIRE FUMF -

Instal *ation 58 a hydroturbine draventAC pcwer andependent) fuwt nit transfer pump to 6 esp the tire pump daytank *ttnas even tilat dur t eg a st ation blackout the diesel fire punc as avail abl e on a cuntanuous basts. The suction anc casche ge lines to the hydroturbine wtli be 1-1/0* an Otameter and will be connected to the discharge and suction ptptng of tha owastan3 diesel fare pump. (Fleed implementation scheduled

! te buyth en July 3, 199/>

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J,UL 06 '87 13147 HRC KING OF PRUSSIA PO4 FLANT CESIGN CHANGE No. DESCRIPfl0N 96-57 BACb, UP NITROGEN SUFF LY (SEP) -

Thta modaftftcation providow two new non-safety related recundant nitrogen sources wh4ch will be available during a station blackout.

The new nitrogen cources back up the e=4 sting containment teertirg system, and the eMtsting drywell instement supply.

The back up nitrogen supply consists of a Inquis nitrogen vaportaer tr ai l er and tec banks (10 cylinders each) of nitrogen cylinders and .nssectated pipteg and val ves, trield l epl ewwot e t t on o igoing)

• 06-56 tutFI ENERGENCY DIESEL SENFRATOR SITE AND CIVIL WORK -

Installation of the thir# diesel generat.or *non-Q) to provide an additional source of back up el ectri cal power and repuce the probability of a complete station blackout.

(Field implementation on hold per State EPA permit) ,

Ob-76 EBG1 MODIFICATION - The modtitcaten includes replaceeent of four automatlc deluge valves with four manual locked closed valves and replacement of the outstang heater circuits with sa f et y-r el a t ed circtuts to preclude single estave fattures.

A single active failure in the SD3T deluge system or an the SSGT heater system can reduce the ef flet ency of the charcoal filter to adsorb radio tedines to the entent that 10CFRIGO.11 !!mits may be exceeded.

(Fleid t mpl ement ati on ongoing) 06-73 auTCMATIC DEPRESSURIZATION/ DLOWDOWN SYSTEM LOGIC (SEPs -

This modt f t cation tncludes a timed bypass (11 min.) of the high drywo!! pressure int tn ation signal and a manual inhtb6t .

of est eting ADS actuatton logic tn provide automatte ADS initiation, If r equ i r es , f or event s such as a break enternal ,

to the drywell or a stuck open sRV. Manual inhibit santches allow the operetne to inhibit ADS operation wattset ,

repeatedly pressing the reset pushbuttons.

(NUREG 0737- !!.K.3.lg) -Field toplementation ongoing l

U6-75 ENk!CHED BORON M0DIFICA T!DN TO ELCS (SEP) -

1ht s modi fication includes replacement of the existing sodium pentabor at e solution by enriched sodtum pentaborate l celutson teorichment of a ma ng mum Sa.1 4 tom persent MalOn wath*manteum concentr ation of 0.42 weight percent. The SLCS. using the ==tetang concentration of sodt ura rant abor a t e a

sol ut i on and wt t t. tne eetsting pump capactty. Oces not meet the NRC wfWS rule (4QCFR50.eC) reautrements of mtntmum fica capacity and Ocron content equivalent in control capartty of 86 som of 43 wenght percent of sodism pentaborate.

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.hUL.06'0713:48 NRC KING OF PRUSSIA P05 e . =1 FLAN 1 DESIGN CHANGE NO. DESCRIPTION So-BO REACTOR VESSEL HYDROGEN WATER CHEMISTRY ( Ht.C ) FROSE -

This modif1Catton includes the removal of a SRM, the nnstallation of an in-core HWC probe in place of the SRM, the removal of the NWO probe et the completion of this fuel cycle (c yc l e.9) , and reinstallatten of a SRM. Since data on in-core chemistry s 11matee, *.he actual margan agannst streer; corroston c'aacking of highly treadiated sumponente af f orded by hvorogen water chnet ett y is unkunn.

Electrochemical potential testing via HWC probe will pr wide en assessment of that margin. (Fleld i mpl ement ati on ongoing) 86-10 FEEDWATER PUNP Tnl/ MODIFICATION (SEP)

Thns modtitcation provides and A1WS trip of the feedwater pumps by installing a new trip coil within the breaker 1he curren'. ATWS associ ated wi tle wach resetse # sed pump.

design cunstwL uf telpe of the rect rculation pumpe and inttnation uf tiew automette red insertson (AR!) system on low low water level or high reactor pressure. The feedwathe pumps will be tripped on high reactor pressure (1400 psig) only. - Fleid t est ementation ongoing 87-30 ATWS RECIRC FUMP TRIF' NODIFICATION (SEP)

This modtitcatton provides an ATWS tens af the r ecirc pumo N/G drive motor breaker to increase the reliability of i RPT.eTW9 The enrrent design trippe the recirculation M/G l l field breaker after receipt of high reactor pressure or l

low reactor water level. The drtve motoe trip will use the entsting inttnating signals for the field breaker trap without the surrent time delay (10 sec.) bef ore int t n ation of the pump telp on low reactor water level.

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CORE RELOAO AT PILGRIM

-September 1987-In anticipation of Boston Edison Company (BEco) couencing operations to reload the nuclear fuel back into the reactor core at Pilgrim in lat,e September 1987, the staff examined existing analyses of a refueling accident and considered possible of fsito consequences of such an accident.

This examination was prompted, in part, by concerns regarding the adequacy of offsite emergency preparedness in the area surrounding the pilgrim Station.

' The effects of a fuel handling accident are analyzed in Section 14.5 and in Appendix R of the Pilgrim Final Safety Analysis Report. Portions of those analyses are sum.marized herein.

Accidents that result in the release of radioactive materials directly to the containment can occur when the drywell is open. A survey of the various conditions that could exist when the drywell is open reveals that the greatest potential for the release of radioactive material occurs when the drywell head and reactor vessel head have been removed. In this case, radioactive material released as a result of fuel failure is available for transport directly to the containment.

Various mechanisms for fuel failure under this condition have been investigated. With the current fuel design the refueling interlocks, which impose restrictions on the rovement of refueling equipment and control rods, i prevent an inadvertent criticality during refueling operations. It is concluded that the only accidant that could result in the release of significant quantities of fission products to the centainment during this mode of operation is one resulting from the accidental dropping of a fuel bundle onto the top of the core.

The sequence of events applicable to this type of fuel handling accident are described below:

1. The fuel asse-bly is dropped from the manimum height alici.ed by the fue? nandling equipment.

2. The entire atcunt of potential energy (referenced to the top of the reactor core) is available for application to the fuel assecblies involved in the accident. This assumption neglects the dissipation of some of the mechanical energy of the falling fuel assembly in the water above the core and requires the corrplete detachment of the assembly from the fuel hoisting equipment. This is only possible if the fuel assembly handle, the fuel grapple, or the grapple cable breaks, or improper grapplings occur.

3. None of the energy associated with the dropped fuel assembly is absorbed by the fuel material (uranium dioxide).

The FSAR provides a conservative analysis of postulated fuel rod f ailures, fission product releases, and radiological consequences of the fuel handling accident. The FSAR analysis assumes the fuel he s received an average irradiation time of 1000 days at desir- power up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the accident. (In the present case Pilgrim has been shut down for about 1 1/2 years. Consequently the fuel has not been exposeo to irradiation at power for this period).

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.g. l The following assumptions ar.d initial conditions are us*a in the FSAR I calculation of the fission product release to the environs, t 1 Hitt radiation levels in the Reactor Building refueling ventilation )

exhaust will isolate the normal ventilation system and act W ) the l Standby Gas Treatment System. (Section 3.7.8 of the Technical .

Specifications provides requirnments for Stardby Gas Treatment System i operability during fuel handitag operation). {

2. Since the refueling accident does not result in the release of any liquid or vapor to the secondary containment 3Abe normal environmental condition existing prior to the accident will also exist af ter the accident, except for the addition of the released f fission products. The relative humidity in the secondary l containment will therefore be considerably below any levels which may be detrimental to the filter media in the Standby Gas treatment System. However, the air flowing through the filter system is heated to reduce the relative humidity to 70 percent or less.

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3. The filte- efficiency is assumed to be 99 percent for todines and 0  !

percent for the noble gases. I l

4. There is one secondary containment air change / day through the i Standby Gas Treatment $ystem.

l The FSAR provides tabulated activities espected in the containment and f postulated fission product activity released to the environment. The i radiological exposures to the general population have been evaluated for sin  !

meteorological conditions ranging from very stable to unstable meteorology (

occurring with ! m/s and 5 m/s winds. Two exposure periods have eeen  ;

evaluated, a 2 hr. exposure period and a 24 hr. esposure period, comeonly efereed to as the total dose, it should be empharized that the radiological exposures present in the FSAR are based upon the assumption that the stated meteorological conditions entst for the duration under consideration and that p the wind blows 1a one directica during the entire release period.

Based upon conservative estimate of a release and atmospheric conditions, the 24-hour whole body and thyroid inhalation doses provided in the F5AR are f 2,1 m 10 -2 rem and 1.0 x 10'I res respectively for an individual loca*ed 2,200 meters $5W of the main stack (Manomet Hill). Ther,e doses, are below the guideline whole body and lifetime thyroid doses of 25 rem and 300 rem (10CFR100),

respectively.

I If consideration is given to a 30 day dese, the 24-hour cloud gamma values would be increased by a factor of 1.08 while the thyroid inhalation dose would be increased by a factor of 1.22.

The FSAR analysis concludes that this accident will not result in any i

radiological exposures which endanger the health and safety of the public.

As stated above the 24-hour whole body and thyroid inhalation doses would be less than 2.1 x 10'I rem, respectively. These doses are well below the lower

• range of the Environmental Protection Agency (EPA) Protective Action Guides  !

(PAGs) st which offsite!'cotection actions should be considered (1 rem whole {*

body and 5 ree thyroid).

3 It should be noted that the foregoing FSAR analysis is based on initial core fuel. For reload fuel (which is the present case) the FSAR analysis determines that the radiological consequences will be below those consequences for init?al core fuel.

Consequently, we can conclude that a refueling accident at Pilgrim is conservatively bourded by the FSAR analysis and EPA Protective Action Guides.

Movement of nuclear fuel back into the Pilgrim relctor vessel does not imply that the NRC has made a decision to allow restart of the Pilgrim facility.

Reloading the fuel allows the lit *cittto conduct certain plant operations necessary to demonstrate that the faci W 'v~ ready for restart. The NRC will not permit restart of the Pilgrim faci ity until all the issues raised during the facility's prolonged shutdown have been dealt with to the satisfaction of the NRC.

This summary has been prepared by the Ptigrim Project Manager and coordinated with the staf f members of Region ! and the NRR Division of Emergency Preparedness.

Project Manager: Richard H. Wessman l Dated: September 18, 1937 l

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Agenda for the September 21, 1987 Pilgrim Restart Panel Meeting

• 1. Freparattons for the Sept. 24 BECo-NRC meeting (NRR)

• 2. Emergency Preparedness issues

-exemption request l -response to Varga 8/18 letter

3. Refueling Activities

• -EP posit 1on paper (NRR)

-Open Item Closure

• -Pefueling Inspection Plan

4. Current Outage Schedule

) -Restart Plan Submittal (0-3 weeks)

5. NRC Timetable

-Inspection Schedule /Recent Results

-Response to 2.206

-Public Meeting

• 6. dllegation Status

7. Escurity Meeting (BECo Request) on 9/30

-Discuss Response to Enforcement Conference

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• = Reference mater?.a1 attached Qw b /

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kjr do/s) wtc MEMORANOUM FOR: L. Shao, NRR E. Kane, RI FROM: S. Varg&., NRR SUBJEC1s INITIAL ASSESSMENT CF PILGRIM SAFETY ENHANCEMENT PROGRAM On July 9, 1987 Boston Edison (BECo) submitted a detailed destription of the Pilgrim Safety Enhancement Program (SEP) t c.

the NRC. (Cepies of this submittel hr.ve been provided to you separatelyt however contents of this submittal are sumnarkred in Enclosure 1). 1ho subm3ttal desceibes herdware changes that BECo has voluntarily elected to impicnent at Pilgelm. SECo states in their submittel that none of the physical platit changos increa%es the probability or concequences of a design basis accident and that all of the changes wt11 result in a reduc tion in the frequency of core melt scenarlos or an improvement in the porformance of the containment response. FECo has advisod t he Pi1 that all of the changes could be impicmented under the provissons of to CFR 50.59.

SECo has advised the staff that they intend to implement those changes prior to restart of the Pilgrim facility (entinated by BECo to be in late September). In a July 16, 1987, convetuation with Ralph Bird (BECo Senior Vice-President - Nuclear), Dr.

Murley commited to a prompt staff assessment of these changos te.

determine their safety impact and to evalisate the licensee's

1. pproach to their implementation. As part of our initial escessment of these changes a visit to the SECo engannering offaces in Braintree MA in planned for Suly EP. 1987.

I have directed the Pilgrim PM (Dic.k Wassman) to lead a miiltidisciplinod team including both NRR and Region I personnel

'to make this visit. Guggestod representatives are idontified in Enclosure 2.

To structure the team's effort and to allow me to report the results of this initial assessment promptly to I)c. flurley and the utt1 sty, the guidelinos and summary report formet of Enclosnies 3 and 4 should be followed. Alto included for information is guidc.nce regarding 16 CFR 50. 59 reviews e>:t rac te d from the IE Manual (Encionure 5).

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I apprer.iate your "short-fused" support on this effort. Pleano contact B. Boger (x27415) or D. Wessm n (H24937) if you have questions. TAC Mo. f=43 5 6 'I '

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Steven A. Varga, Director Division of Reactor Projects, 1/II E nc i cisur e s As stated Copy to:

T. Murley F. Miraglia .

R. S t ar c4 tec k i A. Thedani F. Rosa J. Creig W. Hodges J. Joyce J. Wiggins, R!

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1.0 INTRODUCTION

1.1 Purpose of Report ,

1.2 Scope of Report 1.3 Safety 8:nhancement Program Goals 1.4 Safety Enhancement Program Plant and Operational Changes 2.0 OVERVIEH OF SAFETY ENHANCEMENT PROGRAM

2.1 Background

2.2 Safety Enhancements

3.0 DESCRIPTION

OF SPECIFIC PLANT SAFETY ENHANCEMENTS 3.1 General Considerations 3.2 Installation of a 01 rect Torus Vent System (OTVS) 3.3 Containment Spray Header Nozzles -

3.4 Additional Sources of Water for RPV Injection and Containment Spray 3.5 Diesel Fire Purrp for RPV Injection and Containment Spray 3.6. Diesel Pump Fire Pump Fuel Oil Transfer System 3.7 Backup Nitrogen S[ijipl~ySystem 3.8 Blackout Diesel Generator Including Protected Installe. tion Facilities . .- .

3.9 Automatic Depressurization System Logic Modifications 3.10 Addition of EnH eh'ed Bo'ron 'to'Stan'dbyL161'd Cont'rd1"Sys tem 3.11 ATHS Feedwater Pung Trip 3.12 Hodifications to Reactor Core Isolation Cooling System Turbine Exhaust Trip Setpoint , , , , , , , . . , _ .

3.13 Additional ATHS Recirculation Pump Trip

4.0 DESCRIPTION

OF OPERATIONAL PLANT SAFETY ENHANCEMENTS *

5.0 CONCLUSION

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D 20 87 ' 07:15 to.002 005 --

Enclosure 2 I N I T I AL ASSESSMENT TEAM MEMBERS i i

FILn_c t_i c.n* L Ar e a OI.R.ad r etio n __ Member.

Menagement/cocidination PD I-3/NRR R. Wwssmas.

Plant Systems 5FL5/NRR C. Tinkler Roactor Systems IR*5/NRR ,

I ns t rumrent a t 1 on SICE/NR9 HLS/NRR Electrical Implementation /50.59 applic. DRS/RI L. Briggs e e e 0 ee e . - *e e

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i Enclosure 3 1NITI AL ASSESMENT _ GlJ1Qggfjgys, The following are suggested guidel snes fc r use in conduct of the initial assessment. Any conclusions reached about the technical adequacy or method of implementa+. ion of the PECo SEP modifications are considered tentative. This assessment is briof and cannot reflect an indepth technical review, due to the constraints of time. Each SEP enhancement should be assessed With consideration of the following:

1. What is the safety impact of the change when considarad alonc or along with the other changes? Does an "unreviewnd sefety question" exist? (Criteria for determining whether an unenviewed safety question exists are defined in Faragraph (a)(2) of to CFR 50.59. Copy attached).

2. Is a change to the Technteal Spectftestions required? (If the answer is "Yes" the modificatic.n and the proposed Technical Specification change must be reviewed by the staff before implementation).

3. For those items in which no Technical Dpecif)ca tion chauge is involved, should the licorsee be allowed to amplement the change before staff review is complete?

4 Assess the adequacy of the licensee's evaluation and

~ conclusions regarding each SEP item.

3?/20 87 07:16 tO.CO2 007 Enclosure 4 EU11 MARY _ REPQRT FORMAT For each SEP item you assess, provide a brief summary repor t, using the #ollowing format. This is to fact 11 tate management dec1sionmaking and assure consistency in approach. Hcipef u l 1 y ,

each summary won't require more than 1-2 pages.

1. Summari:e the propocod SEP item.

2. Summarize your conclusions regarding each of the 1tcmz ti.

Enclosure 3.

3. Provido your recommendations for further NRC ait t ion. c r indicate if you believe no further action ic warranted (othee than routanti inspection of the modification. as ele.ctod by Rnq1oa I).

4. Provide any additional comments you feel are appropikato.

I l

l l_

SEP a7 g ,./.4_Z _, /

,(

Robert F. Janecek, Chairman -

UWR Owners' Group c/o Comonwealth Edison Company

• y Rm 34 FN East P. O. Box 767 g7.L i A - g.em % ./

m%.w ' ' ['^

• r

,/ dr

,;)qf Chicago,IL 60690

_! .i.i u.) c,d x et /

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Dear Mr. Janecek:

S h M. r .

M\.C

Subject:

Containment Venting f> d

'M("Nk The purpose of this letter is to thank you 10,and 1987 the Owners' Group for g'g theregard g excellent venting. presentation you gave on SeptemberYou and your consultants k u;.Vcr4.

greatly helped the staff to understand better the concept g of co venting.

S (Enclosure-1) q.C In the process of further studying this matter, we have identified severaA list of

'e ,

questions concerning containment venting.related to this matter isd en these concerns as soon as po ible.

W To facilitate such discussions, I propose that your responses be provided to Please let me know your views as to a tentative Mius R9 s >

in advance of a meeting.

D re schedule that we could work toward. C ll, Sincerely, km Ortainal Sign +d by MeI R. W. Starwteeli j

hvf Richard W. Starostecki, Associate Director for Inspection & Technical Assistance Office of Nuclear Reactor Regulation

Enclosures:

1. Minutes of the Meeting

2. Questions Concerning the Containment Venting .

Issue O!STRIBUTION A. Thadant Central file M. W. Hodges SRXB R/F R. Jones R. Starostecki G. Thomas T. Murley G. Thomas R/F J. Sniezek L. Shao r ,/I, _

• SFJ PREVICOS COtCURREtKE PAGE :ADI)  : :

D: DEST

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ATHADANI :LSHA0 .- .----:---- -

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.....:.......... .:.........-..:....-.....--:...--.....--:..---- :9/S / 87 -.---:- - a--

9/17/87 :9/17/87 :9/n/87 '

• ) ATE :9/17/87 :9/17/87 OFFICIAL RECORD COPY I f,/[

19~)A M 0 i Of 7{f

~

% UNITED STATES

• t a NUCLEAR REGULATORY COMMISSION w Assiwof oN,0. C. 206S5 l

5

\..../ EhCLOSURE 1 MEMORANDUM FOR: M. W. Hodges, Chief Reactor Systems Branch Division of Engineering & Syste:ns Technology THRU: Robert C. Jones, Section Chief Thennal-Hydraulic Perfonnance Section Reactor Systems Branch Division of Engineering & Systems Technology FROM: George Thomas, Nuclear Engineer Reactor Systems Branch Division of Engineering & Systems Technology

SUMMARY

OF MEETlhG WITH BWR OWNERS' GROUP ON CONTAINMENT VENTING I. General On September 10, 1987, the NRC staft met with representatives of the BWR Owners' Group on Emergency Procedures Guidelines to discuss the basis and logic for BWR contaircent venting. The treeting agenda and the slides presented at the meeting are included in Attachrent-1 to this memorandum.

Attachment-2 is a list of meeting attendees.

The treeting started with preliminary introductory remarks. R. Starostecki j expressed concerns regarding the advisability of containment venting in general 1 and, in particular, early containment venting ~as includeML. rey-LoLthe

~

l Emergency Procedure GiRdeTE 'RfilDhere follcwed a candid presentation 7Cdiscussion of the need for containment venting.

11. Meeting Highlights ,

(1) Robert Janecek (BWR Owner's Group Chairfr.an) opened wit 3 an explanation of how venting assists in prevention of core damage. He also (,'escribed how venting mitigates the consequences of core damage. He noted that containment (

_ venting was first identified in EPGs Rev-0 (1980). (

(2) T. Rogers of Operations Eng, Inc., consultant to the Owndrs' Group, gave a brief surmary of the functional criteria which are used for the EPGs. The EPGs must provide operator response on recognition of particuYar symptoms independent of the causal event. Yhe EPGs must specify appropriate actions .

for all physically possible plant conditions irrespective of probability of

[

t occurrence or specific event sequence. The EPGs must specify appropriate actions for use of all useful major plant equiptnent, irrespective of The its EPGs qualification, safety classification or original intended function.

1 i

I mm

o M.'Wo Hodges ,

should not require actions which are beyond the capabilities of plent systems or operators. The EPG also must specify appropriate actions as currently built irrespective of rodifications which may be incorporated at a later date.

He described the controlling Rogers explained the basic strategy of the EPGs.

parameters used in reactor pressure vessel (RPV) control, primary The containmen control, secondary containment control and radioactivity release The control.

EPGs EPGs identify appropriate actions and indicate limits in advance.

provide parameters.

a graduated response keyed to certain important plant o The EPGs maximize the time

_e.ither core or containrentJs_long_as_possible.The EPGs also include operator actions aviDa61e to recover systems.

for severe accident mitigation.

Rogers described the basic primary containment control sequences used in EPGs and highlighted the sequence of operatnr actions and unavailability of systems required before venting is permitted. suppression pool c shutdown cooling and use of main condenser. A.so the unavailable or insufficient before containment venting is coerenced.

reactor pressure vessel must be depressurized and the primary containment .

pressure must be in excess of that calculated for any design basis acciden Rogers presented the new definition of primary containme (PCPL).

lowest of (1) the pressure capability of the contairment or (2) the maximum contaircent pressure at which vent valves can be opened and closed to reject all decay heat from the contairment or (3) the maximum containment pressur which SPVs can be opened and will remain opened or (4) the maximum centa pressure at which vent Yalves can be opened and closed to vent th primary contaircent ficoding.

water level and temperature, and the limit is used to preclude containment tailure and core damage.

He showed Rogers described the containeent structural failure sequenc He also accident sequences ultimately leading to containment failure.

presented the considerations in selecting a vent path. motlf ~venting is done through a path whi ILp_ asses through the_ pool._then_tbe benefit 1 ac eved.

__ scrubbing;fping pressure p ratfieMarIthe dWtworGWh~ can operste f W to~ only-at-aTery ground lo Use of a

_ pressure is prefierred-for-tent 7ath-Elevated

~

The power~r_elHieXFr source single pathway may be better than using rultiple vent paths. Containment isolatio

• for the vent path must also be considered. The vent path the vent valves may be bypassed for containment venting.

location should be decided with due consideration for rad plant personnel.

M. W. Hodges . actions specified in the containment venting procedures also should be considered.

R. E. Henry of Fauske & Associates, Consultant to the Owners' Group, (3) described the benefits of venting identified by severe accident analyses. He stated that the ir.dustry and the NRC studies clearly indicate the benefit of containment venting. He noted the increased risk associated with not venting.

He presented ICCOR and NUREG-1150 results to indicate the benefits of Containment venting. - (Seo Attachment-1 for details)

(4)

R. Starostecki asked whether the utilities looked at the possibility The of upgrading the drywell spray design to de'.ay the The containment He raised a concern of additional risk involved in containment venting.

Owners' answered that the 10COR and NRC results have both indica rJsk _is_ increased with_not_ venting under_ accident conditions. Je raised the1he issue of inadvertent early venting by the op~erator. It is very unitkely the operator (WpecteTto foll5w Starostecki thUlant procedures.that asked why it is not the ope more than one operator in the control room.

feasible to wait, bottle up the containment, and vent only 2-3 times of the containment design pressure).

The Owners' replied that the vent pressure should not be based o structural limit. If based on scee assumptions which may it is not difficult exist to during there is no guarantee that the vent valves can be reclosed, If the release is forecast the exact location of brecch of the containment. Herce the through other than the suppression pool, scrubbing is not possible. Controlled radiological impact will be more serious.

uncontrolled release where_licenseetcan Containment tate _ anticipatory venting is done as a last _ steps _to_preYen

_resort.

cat as troph i c!co ntainme ntf athre .The likelihood of improvement in accide period between reaching the structural pressure limit and containment fai is very small. ,

(S) t. Shao asked whether there is any impact with regard to types ofThe Owner containment like concrete or steel during venting.

the intent of containment venting is to preYent the failure of containment.

So there is no difference whether it is a concrete or steel typeThe ofOwner's He asked the typical vent pressure currently used.

containment.

replied that the vent pressure varies frcm plant to plant and 70 psig can taken as a typical value.

R. Barrett questioned the decision to vent irrespective of off-site dose ,

(6) He steted that Emergency Evacuation and Accident Management factors limits.

should be considered during an accident scenario which may require contain:

venting. The Owner's replied that during an emergency, there is no guarantee

P. W. Hodges ,

that technical support people or accident management people will be available l at the plant-site at all times to make the decision with regard to off-site dose limits. The A ners'_pos':Jon is_that the._ operators wh; are at the site.should-be.given guidanc, 3o cope wTih any emergency situation shich may nest of the sequences.whichlmay_?ballenge_Qe conta the corttrolledJ. eating will enable the' licensee't6 take necessary

~496-hrs.;EmergencyEvacuation.

steps f6r The C%ners' Group did not discuss containment venting used for (7) combustible gas control.

(8)

R. Starostecki concluded the meeting and thanked the Owners for their presentation.

T

/l C n d 5 I n -j George Thomas', Nuclear Engineer Themal-Hydraulic Performance Section Reactor Systems Branch Division of Engineering & Systems Technology i

u cy.m .

. ENCLOSURE-2 00EST10NS ON CONTAINMENT VENTING 1)

PROVIDE COMPREHENSIVE ANALYSES OF ACCIDENT SEQUENCES, WITH T ESTIMATED FREQUENCY OF OCCURRENCE, FOR WHICH THE VENT WOU UPON TO OPERATE.

f N ~ spel[fc [

2) PROV10E AN ESTIMATE OF THE FRACTION OF THOSE WOULD BE OPERATED BUT WHERE THE ACCIDENT WOU CONSIDER THE FOLLOWING OF CCNTAINMENT FAILURE WITHOUT VENT OPERATION.

SITUATIONS IN THE ACCIDENT SEQUENCES:

(A) ELECTRIC POWER RETURNED TO SERVICE (B) EQUIPPENT RETURNED TO SERVICE (C) HIS-DIAGNOSED SITUATION CORRECTED BY OPERATORS

3) RROVIDE COMPREHENSIVE ANALYSES Of THOSE ACCIDE COULD BE IMPPOVED BY CORPECT USE OF THE YENT, OR (A)

(B) COULD BE INITIATED OR MADE kORSE BY INCOR

4) PROVIVE ANALYSIS OF SEQUENCES THAT COULD OPERATION OF THE VENT FOLLOWED ,-[ "I VM T BY EXCESSIV (BUCXLING). k b T[ $wdh

5) PROVIDE ANALYSIS OF THE PROBABILITY OF VEN

6) PROVIDE ANALYSIS OF MAINTENANCE OR SUPV l THAT COULD INDUCE ACCIDENTS.

7) PROVIDE AN ESTIMATE OF THE RADICACTIVITY THE VENT COULD BE OPENED, INCLUDING BOTH CORRECT U .

PROCEDURES AND INCORRECT USAGE UUE TO HUMAN MALFUNCTION, w

u m syn mo ad &

u rJ, ,% a n ~.A (

2

8) DESCRIBE HOW TO REISOLATE A CONTAthMENT AFTER VEN HOW TO PROCEED IF REISOLATION IS NOT POSSIBLE.

9)

DESCRIBE THE RELAT10NSH7" 0F VENTING TO EMERGENCY PANAGEMEN GUIDANCE TO SURROUNDING POPULACE IF RECLOSURE IS NOT b d.W /- N bd 4ma4ycd A ' ~ t- f .

e

t

• AfASSPIRG MASSACHUSETTS PUBLIC INTEREST RESEARCH GROUP September 30 1987 Dr. Thomas Murley Director. Of fice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission '

Washington, D.C. 20555

Dear Dr. Murley:

I am writing to urge you to investigate severe overti=e problems at the Pilgrim nuclear f acility in Plymouth, MA. Previous SALP reports have made mention of the excessive use of overtime at the plant, and recent accounts underscore the same problem.

Within the last two week s. I have received several calls from employees at the plant and/or th-ir spouses expressing gravc concern about the health of workers and the safety of the plant due to overtime. Unlike the company's state =ents that overtime is under control, these workers explain that "plant workers routinely work 72 and 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> per week for months on end." The enclosed letter shows that overti=e of "personnel responsible for quality assurance, i,nstrumentation and control, radiological health and safety, operations and plant maintenance. . . .results in poor workmanship and a consequent co= promise in safety."

The Feach Bottec experience should bear a lesson on good =ange:ent and the use of overtime. This and other continuing manage:ent, st ruc tural and e=ergency planning issues make the plant unsafe.

It is'with great concern about the health and safety of Pil &rim workers and citi: ens of the state of Massachusetts that : ask you :: pro =ptly investigate this =atter and keep the Pilgri: plant closed unti'. it can be proven that this issue and others have been cc:plete'.y rese'ved. I would also ask that you obtain a copy of the overti=e docu:entation being kept by the outage manage:ent group at the Pilgri: plant.

I would appreciate a response to this request within 30 days.

Thank you for your consideration of this important =atter.

Sincerely.

n '

,y ,

• ~

?

4,, ~

'achel Shi:shak rgy Advocate Enc.

cc: p_fmT *Maini s t ra t o r Region I U.S. NRC 631 Pa rk Ave.

King of Prussia . PA 19406 W.Ebh Yf M Tomnio Placn it n = t n n M n 01111 (A171747.Jonn ' '2D '"

4 e - 3 0 - 0 .'

Tnere is a cr cical problem at the Pilgrim Nuclear Power Station in Flymouth that needs to be addressed, a r.1 it is my hope that af ter reading this, you will do so in your official capacity. It is a hidden problem, and as potential victims of i t t: .' o n t e q u .f r. :s , you and I must bring it out ir : :e Open and f o r'. 'N .' I u t i On .

I am employed at Pilgrim in a management capacity. -

have seen the station evolve over the past 10 . years in terms of upgraded systems and equipment, new procedures, an: a man age-ment commit Ter.t to an improved physical acrkins envir:nment.

I have al.ays considered myself a f e rv e r - suppcrter :f the nu-C1Ca.- 1"."-...*..*". u'-. .

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prescribed by law (yet strongly encouraged by the ::..C) the guide linkeil events or near events at this country 's nu:' ear ;ceer plantn with the excessive amount o f over time require d o f ..;r.< e: s at the piant, ihe guide s trongly sus 6es ted the T.inimiza .1;r.

of personnel overtime , not to exceed 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> per week and six consecutive days.

Several years ago Boston Ecison management was severely c ri ti c i r.cd for understaf fing e.nd for its wholesale ure of : te.-

time. Inp12ci; in this criticistr was the sugges:1cn tha cver-worked and fati6ued personnel c on t ribu :e d '.ar g e'.;  : :.u f2: . ' . -

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of overtime is well documented in two, four-inch thick vo'u.es .

maintaineu by the utiliy 's outage mana6ement 6 r o:s p . Despite their nine, these volumes represent overtime use en'.y since the beginning of 1987.

Diant v. rkers routinely work 72 and 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> per week for montns on end. Consider that many of the werkers have been involved in an outage that han been ongoing for a year or more.

It is reasenable to suggest that those people respor.sible fer the public health and safety have been required to push themselves well beyond the limits of Sood JudCemente and all at the direction of plant ma .ag e n,e n t . Consider, too, the implica ;;r.s of .:rking the equ:valen t vf a week and a h al f t o tw o w e e .: :r a s ; r. ; '. e

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r e c ir.: .' ..; for the maintenance and safe operat;cr. of ', t 31 ; . . . . .

i nee; to emphasa; e that i an neith.tr ang. .;r ..r.... .

but concen.ca, damncd concerm-J. I believe in our in :;; .; i,.

8-30-37 I believo 2n what I do for a living. Many positive strides have been made by Boston Edison with respect to the mo'dernization of our plant. I am saddened, however, that the company has yet to respond positively to itG Sreatest criticism and challenge: the abili';, ' o a f fec tively manage.

i:

plant originally denicned to operate with 250 people, and in ar. era where major outages in similar facilities ty pic ally

.equire 500 tc 600 additional people, why does the Pilgrim site currently have a staf f of over 3400? Why must Boston Ediser, man: sement push its hum.an resourcen to the thresheld of error and beycnd' In vie'a cf all the criticism focused on plc.nt .an a g e r.e n ;

in :ne past, :s this the r.srk of prucent and enlightenec mant,ge-

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.:11 u:repently bring those questions t; .e, and I &: 11 ans er

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8-!C-37 cc: Boston Gisbe The Herald .

The En terprise ,

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