ML20207L042

Provides Results of Suppl Assessment of Safety Enhancement Program,Per 10CFR50.59.Change to RCIC Turbine Exhaust Pressure Trip Setpoint Constitutes Potential Improvement Suggested by Mark-I Improvement Program

ML20207L042
Person / Time
Site:	Pilgrim
Issue date:	10/12/1988
From:	Varga S Office of Nuclear Reactor Regulation
To:	Bird R BOSTON EDISON CO.
References
TAC-65356, NUDOCS 8810170153
Download: ML20207L042 (10)

Text

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• October 12, 1988 3

Docket No.: 50-293  !

Mr. Ralph G. Bird  ;

i Senior Vice President . Nuclear l Boston Edison Company T

! Pilgrim Nuclear Power Station (

, RF0#1, Rocky Hill Road }

l Plymouth, Massachusetts 02360  ;

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Dear Mr. Bird:

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SUBJECT:

SUPPLEMENTAL ASSESSMENT OF THE PILGRIM SAFETY ENHANCEMENT PROGRAM i

Reference:

TAC Number 65356

) On July 8,1987, you submitted a detailed description of the Pilgrim Safety Enhancement Program (SEP) which was being implemented in accordance with the provisions of 10 CFR 50.59. We perfomed our initial assessment to provide an  !

understanding of the SEP modifications and the safety significance of the  :

j changes.

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The results of our initial assessment were provided by letter dated August 21. j

1987. We further requested additional information be provided in that letter >

4 and clarification was also requested in our letter dated June 29, 1988. You (

provided responses to our requests in letters dated February 22, August 18, l j and September 7, 1988.  ;

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We have completed our assessment of all the modifications implemented as part i of the Pilgrim SEP and have provided the results in the enclosure to this letter.  !

j We have determined that the results of the SEP will enhance the overall plant safety and performance of Pilgrim.

l This completes our effort on TAC number 65356.

l Sincerely, Originalsigned by:

i Steven A. Varga, Director I l Division of Peactor Projects I/l!

l Office of Nuclear Reactor Peculatfori

Enclosure:

l As stated j cc w/ enclosure: See next page l

l *See previous concurrence j i nA/

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pNAME:*0NcDonald:rw:*MRushbrook:*RWessman :RB ger (:  :  : : ,

.....:............:............:............:............b... .......:............:........... l l DATE :10/11/88 :10/11/R8 :10/11/88 :100/88

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PILGRIM TAC 65356 SUPP. ASSESS.

DISTRIBUTION:

,DocketlFile NRC & Local PDRs PDI-3 r/f ,

SVarga BBoger MRushbrook DPcDonald OGC EJordan RGrires ACRS(10)

JWiggins, Region 1 l

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Mr. Ralph G. Bird l Boston Edison Company Pilgrim Nuclear Power Station 1 i I

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i 1 Mr. K. L. Highfill Mr. Ralph G. Bird i Station Director Senior Vice President - Nuclear j

! Pilgrim Nuclear Power Station Boston Fdison Company  ;

RFD #1 Rocky Hill Road Pilgrim Nuclear Power Station L l Plyrouth, Massachusetts 0?360 RFD81, Rocky Hill Pead 1

Plymouth, Massachusetts 02360 .

t L l Pesident Inspector's Office Mr. Richard N. Swanson, Manager i

! U. S. Nuclear Regulatory Comission Nuclear Engineering Departnent j Post Office Fox 867 Poston Edison Company  ;

j Plymouth, Massachusetts 0??60 25 Braintree Hill Park l

' Braintree, Massachusetts 02184 i I

i Chairman, Board of Selectmen 11 Lincoln Street Ms. Elaine D. Pobinson i

Plyt.outh, Massachusetts 02360 Nuclear Information Manager  !

i Pilgrim Nuclear Power Station l 1

Office of the Comissioner RFD #1, Rocky Hill Road I

j Massachusetts Department of Plymouth, Massachusetts 023f0 Environmental Ouality Engineering .

One Winter Street Charles V. Barrv (

Poston, Massachusetts 0?108 Secretary of Public Safety  !

r Executive Office of Public Safety ,

Office of the Attorney General One Ashburton Place t One Ashburton Place Boston, Massachusetts 0?l08 i 20th Floor i Boston, Wassachusetts 02108 ,

Mr. Robert M. Hallisey, Director ,

i Radiation Control Progran  !

l Massachusetts Department o'  !

i Public Health l l 150 Tremont Street, 2nd Floor l Boston, Massachusetts 02111  ;

i Reaional Administrator, Region ! l U, S. Nuclear Regulatory Comission  ;

475 Allendale Road l King of Prussia, Pennsylvania 10406  !

j Mr. James D. Keyes l

Regulatory Affairs and Programs Group j j Leader j

Rosten Edison Corrpany I 25 Braintree Hill Park i Braintree, Massachusetts 071P4 I

ENCLOSURE SUPPLEMENTAL ASSESSMEhY : f" .ILGRIM SAFETY ENHANCEMENT PROGDAM

1.0 BACKGROUND

On July 8, 1987, Boston Edison Company (BEco) submitted a detailed description of the Pilgrim Safety Enhancement Program (SEP) to the Nuclear Regulatory Comission (NRC). The NRC staff provided its initial assessment of the SEP in an enclost.re to a letter dated August 21, 1987 to BECo. The letter also requested additional information in the following sections of the initial assessment: the Direct Torus Vent System (DTVS), Section 3.2; the additional sources of water for reactor pressure vessel (RPV) injection and containment spray, Section 3.4; the Backup Nitrogen Supply System, Section 3.71 and, the modification to the Reactor Core Isolation Cooling (RCIC) System turbine exhaust trip setpoint, Section 3.12. BEco responded to the request with the exception of the DTVS, Section 3.2, by letter dated February 22, 1988.

Subsequently, by letter dated June 29, 1988, the NRC staff reauested further clarification on the Backup Nitrngen Supply System. Section 3.7, and the containment spray header nozzle modifications Section 3.3. BEco responded to this request by letter dated September 7, 1988.

A revised design for the DTVS, Section 3.2, was provided by letter dated August 8, 1988. This revision superseded the initial design description provided in the July 1, 1987 letter in its entirety. BECo concluded, on the

,l basis of the revised design, that no changes to the Pilgrim Technical 4 Specifications are required, and they would proceed with the revised design without NRC prior approval.

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] 2.0 Assessment I

The following updated assessments replace the applicable Sections of the initial assessment in the enclosure to our letter dated August 21, 1987.

Section 3.2 - Installation of Direct Torus Vent System (DTVS) -

J The proposed design modification assou ated with the direct torus vent

! system (DTVS) provides a direct vent path from the torus air space to the main stack, in parallel with and bypassing the Standby Gas Treatrient 1 System (SGTS). The DTVS provides a new 8" line branching off the existing torus purge exhaust line between the containment isolation valves (outside containment) with a reconnection to the existina torus purge exhaust line downstream of the SGTS. The new torus vent line is also provided with its own containment isolation valve and a rupture 1 disc. BECo revised the initial design, as detailed in their August 8 j 1988 letter, indicating that the containment isolation concerns had been i

addressed and no changes to the Pilgrim Technical Specifications are

required. They further indicated they would proceed with the revised i j design without prior hRC approval.

On Parch 7, 1988 Boston Edison Company (RECo) personnel met with the NRC >

staff at the Pilgrim site and provided a tour of SEP modifications and an informal presentation of the quantification of competing risks associated with venting the containment and conclusions drawn frem these results.

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The presentation provided responses to the concerns identified in our initial assessment which was enclosed to our August 21. 1987 letter. The l q information was made available to the residents and documented in our  !

Inspection Report No. 50-293/88-12 dated May 31, 1988. An inspection of i the mechanical and structural design of the DTVS was performed and documented

) in Inspection Report No. 50-293/88-07 dated May 6, 1988. (

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I i) As a normal part of our review of licensee actions taken under the i

! provisions of 10 CFR 50.59, the NRC staff will review BECo's DTVS system

) installation and document the results in a subsequent inspection report, ,

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I Section 3.3 Containment Spray Nozzles l Following the staff's initial assessment of this element of the SEP the  !

) staff pursued this matter to review the adequacy of the reduced drywell J j spray flow under design basis accident conditions. Specifically, the i i staff requested that BEco address the effcet of capped spray nozzles and  !

! reduced drywell spray flow on: 1) calculations establishing the

) environmental cualification envelope. 2) analysis of pool bypass '

capability. 3) termination or limiting of pool chugging loads, and 4) l 1 further reduction of drywell spray flows due to nozzle clogging from rust  :

l build-up. l I

With regard to establishment of the environmental qualification envelope

) BECo has confinted that an analysis demonstrating the acceptability of t the environmental quali'ication envelope has been perfonred verifying the  !;

? adecuacy of th6 reduced drywell . gay flow. Therevisedanglysiswas pegformed for a spectrum of steam line break sizes (.01 ft to 1.0 ft ) at a drywell spray flow of 685 gom. The actual drywell spray flow i rate with the nozzle modification is 720 gpm based on operation of one l

train of RHR. throttlin of the suppression pool return valve and i j operation of the wetwel sprays.

Regarding the effects of reduced drywell spray flow capacity on pool l l bypass capability, the licensee has reported the results of analyses  ;

2 3 indicating an allowable pool bygass leakage area of 0.17 ft as compared v l to the maximum value of 0.13 ft reported in the FSAR. The analysis of 1

allowable bypass leakage did not consider the operation of drywell sprays i thus any change in flow rate would not affect the detemination of a maximum allowable leak area.

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Because drywell sprays are used during drywell pipe break accidents to l l reduce drywell pressure and reduce chugging loads, the staff requested the  !

{ licensee address the effects of the nozzle modifications on definition of l

chugging loads. In their response to this matter BECo noted that 5 NE00 21888 Rev. E(November 1982). "Mark ! Containrent Program Load

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j Definition Report" provided the criteria for acceptable chugging duration (

i and that for both intertnediate and small break accidents a chugging l 1

duration of 900 seconds was specified. RECo further responded that i

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3 chugging durations with the reduced spray flow rate are below the specified value of 900 seconds and within the original Mark I load definitions, lastly, the staff noted that in June 1987, BEco discovered that rust was plugging drywell spray nozzles in both the upper and lower spray headers. Because the drywell spray nozzle modification dramatically reduces the number of spray corrles the revised design is inherently more sensitive to the adverse effects of clogging. Therefore. BECo should i provide additional surveillance to assure that corrective actions have been successful in addressing the problem of rust buildup in the spray header and potential nozzle clogging, in response to this issue. BEco

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reiterated that water leakage past the drywell spray isolction valves  ;

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into the spray header durino PHR surveillances was determined to be the )

root cause of rust buildup and nozzle clogging, BECo further noted that I a 3/4 inch drain, designed to remove any water introduced during

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surveillance tasting, has been permanently installed, The drain will be

inspected by RECo for operability during the next refueling outage, i BECo has also incorporated additional surveillance to assure there is no

loose rust buildup in the spray headers. As part of the procedure impleTenting the five year air test of the spray system the licensee has added the requirement for a boroscopic examination of each spray header.

BEco has also incorporated into plant procedures the requirement for a boroscopic examination during the next refueling outage (RF0 #8). The acceptance criterion for the visual boroscopic examination requires that no loose rust be observed in the drywell spray headers, If af ter conducting the next examination there is no evidence of rust buildup or nor le clogoing, then the licensee will begin surveillance on a 5 year l

schedule. The staff has considered the additional surveillance

incorporated by the licensee and finds the scope and frequency to be l adequate, t

i The staff concledes that the lice 9see has adequately considered the  ;

] technical issues germaine to modification of the spray nozzles. l i

Section 3.4 Additional Sources of Vater for RPV Infection and rentainment 5 pray, I

l The basic objective of this design change is to provide additional  !

sources of water that are not dependent on AC power and thus available I for core cooling ard containment spray during severe accidents including  !

station blackout. The design modification consists of a piping crosstte  !

! between the fire Protection System and the RHR system as well as the '

reinstallation of the Rpy head spray line. The RPV head spray line was

{ included in the original design but was disconnected due to water hamer  :

concerns. Reinstallation of the line is accompanied by design changes.  !

rerouted pipino, and a bypass line with restriction orifices added in l order to reduce the potential fer water hamer, j i

The connection between the fire protection system and the RHR system is )

rude by adding a Diping connection from the fire protection system piping r i

ded the RHR Salt Service Water injection line. The design of the  !

connection leaves the path interrupted; when the connection is desired a i removable pipe section. 16" in length, must be installed with quick i

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t connect Victaulic couplings. When the removable pipe section is not i installed the piping ends are capped. Isolation of the RHR system is i provided by the addition of a gate valve (local mani.al) and check valve. l During operation of the RHR system, these valves become part of the '

reactor coolant pressure boundary. Isolation of the line from the fire i protection system is provided by gate valve. The gate valves will be I locked closed. The crosstie on the RHR side of the removable pipe i section is to be designed with ASME Section !!! Class !! piping and ASME l Section !!! Class ! valves (gate valve and check valve). On the fire i protection side of the connection the crosstie is designed to ANS! and r NFPA Standards and is designated Quality Class FPQ (Fire Protection). l r

The effect of these changes allow the use of diesel fire pumps, including i a newly proposed diesel fire pump, which draw water from the fire water j storage tank and the city water supply line to provide water for core injection and containment sprays. In the event of a station blackout, the j non safety related station blackout diesel will be utilized to operate the r POVs in the RHR system for PPV water injection and containment spray  !

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Reference:

Memo from F. Rosa to R. Wessman, dated A ly 31, 1987 "Evaluation i of Station Blackout Diesel Added as Part of Pilgrim SEP"). l i

BECo has evaluated the effect of the proposed design modifications and i concluded that there is no adverse impact en the perforvance of safety }

related systems or the fire protection system. The staff has similarly  !

concluded, that the design changes have no significant deleterious [

effects en the design or operation of the plant. Gate valve 10 F0 511 will be checked for leakage during FNPS procedure 8. A.16. RHR system integrity surveillance, to satisfy the requirements of NUREG 0578. Part  !

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This change will provide an additional source of water to reactor during an i accident and hence will improve the safety of the Ptigrim Nuclear Power  ;

Station.

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There is no known negative impact on any system due to this modification  !

and installation of this system under the provisions of 10 CFR 50.59 i appears acceptable. l Section 3.7 - Backup Nitrocen Supply System l The staff's initial assessment of this element of the SEP indicated that the licensee needed to reassess the containment isolation function l of valve A0-4356. The need for reassessment was detemined based on the e fact that valve A0-4356 was listed in FSAR Table 5.2 5 as a containment  !

isolation valve and that the failed position of the valve. as a result of the SEP modification, was altered from fall Closed to fail open.

BFCo has clarified the original design function of valve A0 43!$ and check valve 31-CK-167 and the containment isolation provisions for the nitrogen supply line penetration. RECo has further indicated that FSAR Table 5.2-5 was in error and that gate valve A0-4356 is not a containment isolation valve. The original design function of this valve which is upstream of check valve 31-CK-167 was to provide remote manual isolation for pneuratic instrument supply line rupture inside containment. This was relevant when instrument air (vs. nitrogen) was the only pneumatic source to operate various valves and instruments. The Backup Nitrogen

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Supply System modification associated with the SEP basically eliminates  ;

concerns over air intrusion into en inerted containment atmosphere since  ;

both the primary and secondary pneumatic sources are now nitrogen.  :

L The containment isolation provisions for the nitrogen supply line i (penetration X-22) consists primarily of check valve 31-CK-167, which is j "ocated immediately outside containment. The check valve is designed to -

Class ! design criteria including seismic qualification. Check valve 31.CK 167 is leak tested as a containment isolation valve in accordance l with the requirements of 10 CFR 50 Appendix J. Gate valve A0 4356 which  !

is located just upstream of the check valve, while not specifically i designated a containment isolation valve, provides redundant isolation  !

capability. The valve was procured as an ASME !!!, Class 2 valve, the I valve operator, however. has not been qualified to Seismic Category I j criteria. Yalve A0-4356 is powered by a panel supplied by swtag bus 06 i which in turn can be powered by either emergency diesel generator. This will provide reliability of AC power in the event a single diesel generator is lost. Valve A0 4356 is a normally open valve which if  !

recessary can be closed by a remote manual control switch located in the i control room. l t

With regard to tho alteration of the failed position of valve A0-4356,  !

BEco has noted that the thrust of this SEp modification is to erovide a  !

reliable source of nitrogen to operate ADS valves for beyond design basis  !

accidents such as station blackout. For that reason A0-4356 was modifteo {

to fail open to provide continued availability of nitrogen to the ADS in [

the event of a loss of AC power. Thus the position of greater safety is l the open position.

l' The staff has considered the rationale offered by the licensee to justify the isolation capability of the n1Mogen supply line and concludes that adeouate containment isolation capability is provided. Redundant ,

capability in the form of check valve 31 CK-167 and gate valve A0-4356 is f sufficient to assure containtrent isolation in the unlikely event the j nitrogen supply. which is needed following an accident, needs to be i terminated. Further the staff concludes that the SEP modification does  ;

not involve a change to the Technical Specifications or o'.herwise involve  ;

an unreviewed safety question. Installation of this mos fication under j the provisions of 10 CFR 50.59 appears acceptable.  ;

Section 3.12 Modifications to teactor Core isolation Cooling System Turbine Exnaust Trip 5etpoint During station blackout ($60) events, the RCIC system is available to l supply cooling water to the reactor and maintain the reactor water l 1evel. The RCIC pump is driven by a turbine using the primary system {

steam. The turbine exhaust is piped to the suppression pool. Continuous  !

discharge of the steam to the suppression pool, however. wJ11 increase i the suppression pool temperature and the containment pressure. The existing RCIC exhaust trip pressure is 25 psig, which will be reached at about 8.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into the 580 event. To extend the use of the RCIC system, the licensee proposed to increase the trip pressure to 46 psig.

This increase of trip pressure will allow the RCIC system to operate '

untti about 15.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into the event, f I

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This change is proposed to enhance the ability of Pilgrim Station to respond to an 580 event and is an instrumentation setpoint change. The design change 4 creases the setpoint of the RCIC turbine exhaust line h:gh pressure ultches, PS 1360-26A/B, so that RCIC can be available at higher suppression pool pressures.

Pressure switches (PS-1360-26A/B) are located in the turbine exhaust piping to detect high exhaust pressure indicative of restricted or blocked exhaust pipiog. These switches eqt to trip the turbine when their setpoint is reached. The present sstpoint of 25 psig was selected to be as low as possible to detect blockage of the exhaust line without causing spurious trips on turbine starts and is consistent with the maximum discharge capability of the gland seal condenser vacuum pump.

The exhaust line and turbine casing are protected against overpressure by a rupture disk set at 150 psi. The design pressure of the turbine casing  !

is 165 psig.

Components directly 3ffected are pressure switches PS-1360-26A/B. These will be re-calibrated to actuate at 46 psig. This value is well within their adjustable range of 5-150 psig. The instrument effects of increasing the preslire switch setpoint have been evaluated and it was con:1uded that the ww setpoint falls below the analysis limit (and the process safety limit) for exhaust pressure even when instrument inaccuracies are taken into account.

The gland seal condenser vacuum pump will not be functional if the turbine exhaust pressure is near 50 psig, as the vacuum pump maximum discharge pressure is 25 osig. The total gland steam leakage, approximately 220 lb/hr fnr 50 psig, will still be condensed in the barometric condenser (partoftheglandsealcondensersubsystem)providedthattheCondensate Storage Tank (CST) is used as a suction source for RCIC. Procadures developed from the Emergency Procedure Guidelines (EPG) Rev-4 direct the '

nperators to align RCIC to take suction from the CST. The barometric 4

condenser, if cooled by CST water, is capable of condensing the entire amount of gland steam leakage.

An increase in turbine exhaust pressure up to 46 psig is expected to cause the turbine control valve to open further to provide for a given i

power demand. Also, the turbine will deliver less power at very low inlet steam pressures when the exhaust pressure is higher. The pump / turbine performance curves published by the turbine vendor was cased on an exhaust line pressure of 50 psig. Therefore, RCIC system capacity and flow rate will still be adequate during those accidents or transients in which RCIC response u part of the current Pilgrim Nuclear Power Station licensing basis. -

RCIC turbine exhaust steam discharge into the suppression pool, where the steam is condensed, results in thermal-hydraulic loads both on the containment structures and the discharge pipe. These loads will be increased significantly with increasing exhaust back pressure. Assessment of the magnitude of these loads is required in order to ensure that the RCIC exhaust pipe will not fail near the increased trip setpoint. Discussions with the BECo's technical staff indicated that the licensee has assessed the loads on the basis of static pressure.

Since experirents and analytical methods indicate that the dynamic load differs substantially from static load, BECo's method based on static pressure a s not acceptable.

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The staff required BECo to conduct an assessment of hydrodynamic loads on the RCIC piping and supports based on the proposed exhaust pressure of 46 psig. The pipe stresses and support / penetration loads for the exhaust piping were previously evaluated in the MARK-I containment long term program for the combined loads that occurred during the simultaneous application of condensation oscillation (CO) shell loading, C0 drag, loads applied to the submerged piping, SEE loads, thermal loads, and weight loads. Since the RCIC turbine has approximately a 10 second start-up time, the gradual start-up will not produce high air clearing loads o.' dynamic effects. Also, steam flow rates through the exhaust line are low. Data from SRV testing indicated that low steam flow rates do not result in significant containment loads. The proposed backpressure setpoint of 46 psig will not result in excessive pressure stress on the turbine exhaust piping. Original design conditions for the RCIC exhaust piping are 100 psig, 325'F. These design values are used in allowable stress analysis and nozzle loads. Therefore, BEco's assessment of hydrodynamic loads on the exhaust piping based on 46 psig is acceptable.

The PCIC turbine exhaust pressure trip setpoint change is one of the potential improvements being suggested under the MARK-I improvement prog ram. 580 is a major contributor to the total risk associated with plant operation. During 580 RCIC with the increased turbine exhaust pressure trip setpoint will be available for a longer period of time.

(RCIC operation is independent of AC power.) Hence, this change will improve the safety of the Pilgrim Nuclear Power Station, and there is no known negative impact on any system due to this modification. Installation of this modification under the provisions of 10 CFR 50.59 appears acceptable.

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