ML20128K941

Responds to Request for Updated Info on Reactor Vessel Instrumentation Issues at Plant.Info from Insp Conducted in Dec 1992 Encl

ML20128K941
Person / Time
Site:	Pilgrim
Issue date:	02/04/1993
From:	Linville J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	Blanch P AFFILIATION NOT ASSIGNED
References
NUDOCS 9302190050
Download: ML20128K941 (2)

Text

_ _ - _ _ _ _ _ _ _ _ ._. _ - . ._. _.

A r

FEB 041993 ,

Docket No. 50-293 Mr. Paul hi. tilanch 135 flyde Road West llartford, Connecticut 06117

Dear Mr.111anch:

In responst u your request for t'.pdated information on reactor vessel instrumentation issues at Pilgrim, enclosed is relevant information from our inspection conducted in December 1992.

This is essentially the same information which was provided to you in our letter of January 7, 1993.

If we can be of further assistance, please contact Mr. IIugene Kelly of my staff at (215) 337-5183.

Sincerely, ,

Original Signed By:

hees C. linyllie t

James C. Linville, Chief Projects Isranch No. 3 Division of Reactor Projects

Enclosure:

NRC Inspection Report 50 293/92 28 (Section 8.3) cc w/ encl:

Ernest C. fladley, Esquire cc w/o encl:

Public Document Room (PDR)

Local Public Documen: Room (LPDR)

Nuclear Safety information Center (NSIC)

NRC Resident inspector 9302190050 930204 POR ADOCK 05000293 G PDR i

I OFFICIAL RECORD COPt L , . _ . . _

i Ium um14 4 Mr. Paul M. Illanch 2 FEB 041993 ,

bec w/o encl:

G. Kelly, DRP J. Linville, DRP J. Macdonald, SRI - Pilgrim x

1:DRP RI:Dy$'

(

Tm(eA L le, 2/2/93 g )

OFFICIAL RECORD COPY A:PilLANCli.MEO

• Y f * * %c UNiiln LT Alls a' f4

[. p g)l#

g> ( e Nucit: AH H[GUl.ATORY COMMISSION m onin g ;e4 y [ 41b tilll NDAl f itOAD s- , ~. o, . sm n ,ms.,, ,m ,m

., .... 7 ,

Docket No. $0 293 2 8 $3 Dr.11. Thomas Ikiulette Acting Senior Vice President-Nuclear lloston Edison Comjuny Pilgrim Nuclear Power Station RFD #1 Rocky liill Road Plymouth, hiassachusetts 02360

Dear Dr. Ikiulette:

SUluliCT: PILGRlhi INSP11CTION 92-28 This refers to the safety inspection conducted by hiessrs. J. hiacdonald, A. Cerne, and D. Kern of this office November 24 to December 31, 1992 at the Pilgrim Nuclear Power Station, Plymouth, hiassachusetts. Areas relevant to the health and safety of the public examined during this inspection are described in the enclosed report. Our findingr, were based upon observations of performance and independent evaluations of safety systems and quality records. The preliminary results have been discussed with hit. L. Schmeling and other members of your staff at the conclusion of the inspection period.

Good precautionary measures were taken during the storm on December 11 14, that included reduced reactor power operations. Your operating staff's response to the weather induced automatic reactor trip was prompt, particularly the shift supervisory oversight of trip recovery activities and electrical distribution system status, llased on the results of this inspection, certain of your activities appeared to be in violation of NRC requirements, as specified in the enclosed Notice of Violation (Notice). It appears that several opportunities existed (but were missed) as part of supenisory procedure review, incident critique, and post-trip report processes to identify the improperly established main steam line (htSL) high radiation alarm setpoint prior to reactor startup. Setting of the alarm was of minor safety significance in that the reactor protection system function is independent of the alarm.

You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response, in accordance with 10 CFR 2.790 of the NRC's

" Rules of Practice,* a copy of this letter and its enclosures will be placed in the NRC Public Document Room. The responses directed by this letter and the enclosed Notice are not subject to the clearance procedures of the Office of hianagement and Budget as required by the Paperwork Reduction Act of 1980, Pub. L. No. 96.511.

l--

JAN 2 8 Y393 Dr. li Thomas Ikmlette 2 Your cooperation with us is appreciated.

Sincerely, g bh/t ames Linville, Ch'-

Projects Branch N . 3 Division of Reactor Projects

Enclosures:

1. Notice of Violation

2. NRC Inspection Report No. 50-293/92-28 cc w/encis:

E. Kraft, Acting Vice President, Nuclear Operations and Station Director L. Schmeling, Plant hianager V. Oheim, Manager, Regulatory Affairs and Emergency Planning Department D. Tarantino, Nuclear Information Manager N. Desmond, Compliance Division Manager R. Itallisey, Department of Public Ilealth, Commonwealth of Massachusetts R. Adams, Department of Labor and Industries, Commonwealth of Massachusetts Tne lionorable Edward M. Kennedy The lionorable John F. Kerry The lionorable Edward J. Markey The lionorable Terese Murray The lionorable Peter V. Forman B. McIntyre, Chairman, Department of Public Utilities Chairman, Plymouth Board of Selectmen Chairman, Duxbury Board of Selectmen Plymouth Civil Defense Director Paul W. Gmmer, Massachusetts Secretary of Energy Resources Sarah Woodhouse, legislative Assistant A. Nogee, MASSPIRG Regional Administrator, FEMA Office of the Commissioner, Mase.husetts Department of Environmental Quality Engineering Office of the Attorney Ocneral, Commonwealth of Massachusetts T. Rapone, Massachusetts Executive Office of Public Safety Chairman, Citizens Urging Responsible Energy .

Public Document Room (PDR) 12 cal Public Document Room (LPDR)

Nuclear Safety Information Center (11 SIC)

K. Abraham. PAO (2 copies)

NRC Resident inspector Commonwealth of Manachusetts, SLO Designee l

U.'S. NUCLEAR REGULATORY COMMISSION.

REGION I Docket No.: 50-293 Report No.: 92-28 Licensec: Boston Edison Company .

800 Boylston Street Boston, Massachusetts 02199 Facility: Pilgrim Nucicar pewer Station laation: Plymauth, Massachusetts Dates: November 24 - December 31,1992 Inspectors: J. Macdonald, Senior Resident inspector A. Cerne, Resident inspector D. Kem, Resident I c.pecto M'  ! [

Approved by:

E. Kelli, C;d f, Reactor Pr ccts Section 3A ' Date Scope: Resident inspection addressed the areas of plant operations, radiological controls, maintenance and surveillance, emergency preparedness, security, safety assessment and quality verification, and engineering and technical support.

Initiatives selected for inspection included: restoiation from an electrical backfeed lineup; observation of an inplant emergency preparedness drill; control and testing of certain containment isolation valves; and, plant design changes associated with the reactor vessel head spray lines, inspections were performed on backshifts during November 30 and December 1 4, 7,11,13-18, and 21-31,1992. " Deep" backshift inspections were performed on December 13 from 10:00 to 12:00 p.m. and December 14 from 00:01 to s 05:45 a.m.

Findings: Inspection results are summarized in the Executive Summary.

Procedure 3.M.2-7.6, "NUMAC Img Radiation Monitor Setpoint Change Procedure" was not properly performed 'lechnicians established incorrect RPS protective setpoints and management reviews failed to identify the associated discrepancies (Violation 92-28-01, see Section 4.4).

The technical basis for the deactivation of a head spray line remains unresolved (Unresolved item 92-28-02, see Section 8.2).

DOde i 1

t - - - - - - _ _ _ - _ _ _ _

EXECUTIVE SUMM ARY Pilgrim luspection Report 50-293/92-28 Plant Operations Operations Section preparation for and response to the effects of a northeaster storm were comprehensive. Decisions to maintain reduced reactor power at the 80% rod pattem line and to separate the safety-related buses from the distribution system deraonstrated a strong safety perspective.

The immediate response by operators to two automatic reactor trips was appropriate.

Communications, use of procedures, and supervisory oversight of control room operations were excellent during post-trip recovery activities and subsequent reactor startups.- Also, the identification of loose or missing bolts on motor operated valve actuator limit switch covers during routine rounds indicated good questioning attitudes and attention to detail by plant .

operators.

Maintenance and Surveillance Actions taken to verify the presence of and trend the effect of-steam leakage past safety relief valve (SR*A RV 203-3A were thorough, Although not required by Technical Specifications (TS), the decilian to establish cold shutdown and replace the leaking SRV pilot valve following an unrelated pw shutdown demonstrated sound safety judgement.

In addition, coordination between the materials & component engineering section, maintenance personnel, and system engineers to complete the repair during this unscheduled maintenance period was outstanding. Restoration from the backfeed electrical lineup following post trip ;

corrective maintenance was performed. Maintenance and operations personnel demonstrated excellent procedural knowledge and communications.

An automatic reactor trip on December 20 was caused by procedural weaknesses and poor work practices by technicians changing the main steam line (MSL) high radiation protective setpoints.

Alsc, the technicians failed to lower the MSL high radiation alarm setpoints following the reactor trip. As a result, the MSL high radiation alarm was not avas.ble to control. room operators upon the subsequent plant restart. Failure to properly reestablish the MSL high -

radiation protective setpoints and associated failure of the management review process on two-occasions indicates a need for greater management attention.

Emergency Preparedness The .apability to draw, analyze, and provide real time post-accident sampling system data under simulated emergency conditions was successfully demonstrated in a December drill, Safety Assessment and Quality Verification Implementation of Phase II of a planned three phase structural reorganization, to become effective January 1,1993, was announced on December 16, 1992. Licensee event reports (LERs) were of good detail, accurate, and clearly identified root cause and corrective action, detailed and properly addressed the required reporting criteria, ii

(EXECUTIVI: SUMM AltY CONTINUED)

Engineering and Technical Support Deactivated head spray line containment isolation valves "

remain to be removed from the Type C local leak rate test program. Several questions regarding -

American Society of Mechanical Engineers (ASME) Code criteria and the technical basis of certain aspects of the head spray line deactivation plant design change remain unresolvedi Continuing NRC review of the licensee reactor vessel water level instnamentation spiking status ,

determined the operability assessment was consistent with the guidance of NRC generic documentation for degraded or nonconforming conditions on operability.

9 4

iii

14 The inspector also checked the compatibility of the Residual Heat Removal (RHR) system configuration, as-left after .nplementation of PDC 86-20 with the design intent of PDC 86-528-196 and, in this regard, additionally reviewed drawings M100BC-282-1, h1100-38-7, hilN 40-12, hi241 and an earlier FRN 191 to PDC 86-52B. During a plant in pection tour, the inspector noted that the electrical supply breaker (B20B3) for the head spray valve (610-1001-63) which had been removed during RFO 8 was still danger-tagged open. The inspector reviewed the PNPS tagout sheet '190-10-21 and determined that this open tag status was inconsistent with the handling of the electrical supply to the other head spray CIV (hiO-1001-60), where the breaker had been left open, but the tagout cleared. The inspector discussed this inconsistency with cognizant plant personnel who initiated action to query the nuclear engineering department as to whether the entire tagout 'I90-10-21 could be closed and cleared.

In review of PDC 86-52B-196 design change criteria, the inspector identined a statement which implied that the abandoned piping in the reactor head spray system outside containment would remain vented. However, since PDC 86-20 installed a pipe cap on one side of this piping and PDC 86-52B-196 capped the piping at penetration X-17 on the other side, the reviewed licensee documentation provided no indication how such venting was implemented in that the valves identiGed in the PDC relating to this piping appeared to have been left in the closed position.

Furthermore, since valve hiO-1001-60, as identified in tagout '190-10-21 was chained closed, two vent paths, one on either side of the closed valve, would have to have been provided for the as-left piping to be consistent with the PDC design criteria.

Additionally, the inspector con 6rmed that the pipe cap for the penetration X-17 piping had been procured to AShiE Code,Section III, Class 2 criteria as " impact-tested material." However, it appeared that an impact-tested weld procedure had not been used to install the pipe cap as would be required by the AShiE Code,Section IX, unless certain conditions of exemption allowed by Section 111 of the code were satis 6ed. Given that the PNPS FSAR documents the containment drywell shell material to be fabricated of impact tested plate and forgings, the licensee issued problem report 93-9005 to resolve this question regarding the weld procedure quali0 cation.

l The inspector determined through the review of PDC 16-52B-196 and related supporting documentation that the current configuration of the reactor head spray piping was acceptable, in that the continued safe operation of PNPS had not been adversely affected by the design l modification. However, as noted above, certain questions, regarding the existing pipe venting i

and the containment penetration weld quali0 cation criteria, remain open. Pending the licensee presentation of evidence that the installed configuration is in compliance with the intended PDC design criteria, these issues remain unresolved (92-28-02).

8.3 Reactor Vessel Water Ixvel Instnamentation Update NRC Inspection Report 50-293/92-23, Section 8.1, provided a detailed status of licensee activities in response to reactor vessel water level instrumentation spiking e perienced during

recent reactor shutdown evolutions. Specifically, the issue has been addressed in terms of the

! generic concern for level instrumentation inaccuracies during rapid depressurimtion events due m the evolution of noncondensible gases from the reference legs.

15 The NRC conducted review of BECo operability determination of the level instrumentation', with particular attention on the instrumentation associated with the two-thirds (2/3) core height containment spray interlock. The safety function of this instrumentation is to provide level-signals and indication such that adequate core cooling can be achieved for certain classes of accidents. The NRC staffindependently concluded that this safety function would be satisfied at Pilgrim based upon the following:

• If flow is diverted to containment spmy after 2/3 core coverage is achieved, one core spray pump alone is adequate to maintain 2/3 level and core cooling. Thus, even the diversion of all available low pressure coolant injection (LPCI) would not preclude adequate core cooling.

• It is unlikely that signincant diversion of flow would occur prior to reflooding the vessel to 2/3 core height, because:

-- the interlock does not cause any automatic actuations; that is, satisfying the interlock does not automatically divert LPCI flow to containment spray.

-- according to the Pilgrim Reload Analysis (S AFER/GESTR Report NEDC-31852),

for desi3n basis loss of coolant accidents, the core is reflooded to 2/3 core height within approximately 60 to 150 seconds; therefore, the operator would have to immediately divert LPCI for such erroneous action to occur prior to reflooding the vessel. Moreover, operators are directed by procedure to assure adequate core cooling prior to initiation of containment spmy, and operators have been sensitized to potendal errors in level indication. Station Emergency Operating Procedures (ie, EOp-03, " Primary Containment Control") which govern the decision to divert LPCI flow and spray the containment would require the presence of a high drywell pressure above 2.5 psig. Also, the EOPs direct that only "those RHR pumps not required to assure adequate core cooling by continuous operation in the LPCI mode" be used for containment spray diversion.

-- over 20 linear feet of reference leg volume, including both horizontal and vertical-sections, must be voided and not recovered at Pilgrim to cause a continuous 14 inch level error, and an error of this amplitude is already considered in the interlock setpoint.

- it is expected that the magnitude of error in the level indication following an actual depressurization event would - be significantly less than that estimated by_

conservative assumptions used in the calculations performed by the General Electric Company and the BECo consultants.

- the potential for level errors has likely been lessened by actions taken by the licensee to reduce external reference leg leakage (ie, tighten fittings and packing at the instrument racks).

• Ifit was postulated that LPCl now was prematurely diverted to containment spray, the safety function of the interlock would still be fulfilled, because:

16 at Pilgrim the diversion of LPCI now to containment spray (both drywell and toms) represents only approximately 25% of the capacity of one LPCI/RHR pump.

- Appendix K analysis from the current Pilgrim Reload Analysis for the most -

limiting case for which LPCI flow is credited (ie, battery failure case), indicate a 1694 degrees F peak clad temperature (PCT), which represents a 506 degrees F margin to the 2200 degrees F PCT limit.

- NRC staff reviewed analysis for a similar BWR/3 plant in which 100% of the flow of one RHR pump was assumed to be diverted from the core from the onset of the accident. These analysis support the staff judgement that, using Apgndix K analysis assumptions, diversion of 25% of the flow of one RHR pump would not result in exceeding 2200 degrees F PCT.

Based on the above, the NRC staff concluded that any manual actuation of containment spray which is based upon an erroneous level signal to this interlock is both highly unlikely and of low safety significance, and that the safety function of the level instrumentation system at Pilgrim would therefore be fulfilled. The NRC staff also concluded that the BECo operability determination was performed consistent with the guidance of NRC Generic letter No. 91-18, "Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability," dated November 7,1991.

During reactor depressurization following the two automatic trips that occurred during this inspection period, no reactor vessel water level instrumentation " spiking" was observed. The reactor tripped on December 13,1992, after 20 days of operation and again on December 20,1992, shortly after startup from the previous trip, The NRC will continue ~to monitor 'BECo's progress in resolving the problem of noncondensible gas accumulation in the level instrumentation system at Pilgrim.

9.0 NRC MANAGEMENT MEETINGS AND OTIIER ACTIVITIES (30702) 9.1 Routine Meetings At periodic intervals during this inspection, meetings were held with senior plant management to discuss licensee activities and areas of concern to the inspectors. At the conclusion of the reporting period, the resident inspector staff conducted an exit n eeting _on January 7,1993 with licensee management, summarizing inspection activity and preliminary findings for this report period. No proprietary information was identified as being included in the report.

9.2 Other NRC Activities During the weeks of November 30 - December 4,1992 and December 14-18,1992 z Probabilistic Risk Assessment team inspection was conducted. Inspection results will be documented in NRC Inspection Report 50-293/92-81.