_ o. .. U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT ' Region I Report No. 50-293/81-18 Docket No. 50-293 License No. DPR-35 Priority -- Category C Licensee: Boston Edison Company M/C Nuclear 800 Boylston Street Boston, Massachusetts 02199 Facility Name: Pilgrim Nuclear Power Station , Inspection at: Plymouth, Massachusetts / Region I Office, King of Prussia, Pa. , Inspection conducted: une 15 - July 10, and September 25-30, 1981 Inspectors:fu /' M ''* $/[8/ /''J . Ja hris o ior Resident Inspector date signed , j . W b |$O - J. McCann lefactor Inspector date signed ~ Approved by: O //k T. El%wfser, Chief, Reactor Projects Section IB, date 'si gned Division of Resident and Project Inspection Inspection Summary: Areas Inspected: Special unannounced inspection of licensee's compliance with 10 CFR 50.44. This inspection involved 20 hours by the resident inspector onsite and 28 hours of in-office review of documentation by one region based inspector. Results: Five apparent items of noncompliance were identified: failure to comply with the provis_ ions of 10 CFR 50.44; failure to perform an adequate 50.59 review; failure to provide appropriate procedures and drawings; failure to make a report requiref by Technical Specifications and failure to provide accurate information to the NRC. These individual items are discussed in this report and have been referred to NRC management for further review and consideration relative to appropriate enforcement action. Rwton I Form 12 (Rev. April 77) , 8202040304 810119 PDR ADOCK 05000293 rf) PDR _ _ .. _ . _. . . _ . _ - _ _ _ .
Transcript
_
o...
U.S. NUCLEAR REGULATORY COMMISSIONOFFICE OF INSPECTION AND ENFORCEMENT
' Region I
Report No. 50-293/81-18
Docket No. 50-293
License No. DPR-35 Priority -- Category C
Licensee: Boston Edison Company M/C Nuclear
800 Boylston Street
Boston, Massachusetts 02199
Facility Name: Pilgrim Nuclear Power Station-
,
Inspection at: Plymouth, Massachusetts / Region I Office, King of Prussia, Pa.,
Inspection conducted: une 15 - July 10, and September 25-30, 1981
Inspectors:fu /' M ''* $/[8//''J . Ja hris o ior Resident Inspector date signed,
j.
W b |$O-
J. McCann lefactor Inspector date signed~
Approved by: O //kT. El%wfser, Chief, Reactor Projects Section IB, date 'si gnedDivision of Resident and Project Inspection
Inspection Summary:Areas Inspected: Special unannounced inspection of licensee's compliance with10 CFR 50.44. This inspection involved 20 hours by the resident inspectoronsite and 28 hours of in-office review of documentation by one region basedinspector.Results: Five apparent items of noncompliance were identified: failure tocomply with the provis_ ions of 10 CFR 50.44; failure to perform an adequate50.59 review; failure to provide appropriate procedures and drawings; failureto make a report requiref by Technical Specifications and failure to provideaccurate information to the NRC. These individual items are discussed inthis report and have been referred to NRC management for further review andconsideration relative to appropriate enforcement action.
The inspector also interviewed other members of the operations, maintenance,and technical staffs.
2. Compliance with 10 CFR 50.44 (Standards for Combustible Gas Control)
On May 29, 1981, as a result of inquiries from the NRC, the licenseeidentified a potential noncompliance with 10 CFR 50.44. /. meeting wasthen held on June 18, 1981 in Bethesda, Md. to discuss present and pastcompliance of the licensee's nitrogen purge system with the requirementsof 10 CFR 50.44. The NRC Offices of Nuclear Reactor Regulation andInspection and Enforcement were represented at the meeting.
It was determined that the Office of Inspection and Enforcement wouldaddress the area of previous noncompliance and that NRR would ensurepresent compliance. In order to determine the extent of past noncompliance,the inspector held discussions with licensee personnel and reviewed thefollowing documents concerning the combustible gas control systems atPilgrim:
BECo letter M. J. Feldman to Director, Division of Reactor Licensing--
NRC letter T. A. Ippolito to G. C. Andognini dated March 14, 1979.--
-- BECo letter #79-114, G. C. Andognini to T. A. Ippolito dated June 6,1979.
-- BECo letter #79-206, G. C. Andognini to T. A. Ippolito dated October19, 1979.
NRC letter T. A. Ippolito to G. C. Andognini dated October 30, 1979.--
-- BECo letter #81-127, A. V. Morisi to T. A. Ippolito dated June 15,1981.
_
...
3
BECO Prompt Telecopy Report (LER 81-021/01X-0) dated June 16, 1981.--
-- NRC letter T. M. Novak to A. V. Morisi dated June 26, 1981.
BECo letter #81-138, R. D. Machon to Region I Director dated June--
29, 1981.
-- Pilgrim Nuclear Power Station Safety Evaluation Report dated August25, 1971.
a. Background
The need for hydrogen control after a design basis loss of coolantaccident is discussed in paragraph 4.1.2, Containment AtmosphereControl, of the Pilgrim Nuclear Power Station Safety EvaluationReport (SER) which states in part, "We have concluded that a hydrogencontrol system should be provided, in addition to the purging systemproposed by the applicant, to keep the hydrogen content withinsafety limits; i.e., less than 4 volume percent...We have concluaedthat the backfitting of the Pilgrim facility in this regard willprovide substantial additional protection required for the publichealth and safety, but that the design and installation of thesystem need not be completed prior to issuance of an operatinglicense. We believe this action to be consistent with the advice ofthe ACRS...that action should be achieved on a reasonable timebasis." Furthermore, the SER concluded that containment inertingwith nitrogen was.necessary in order to mitigate the consequences ofhydrogen evolution following a loss-of-coolant accident. The operatinglicense was issued on June 8, 1972.
The licensee submitted Amendment 35 to their Final Safety Analysis;
Report (FSAR) on January 28, 1974 in response to NRC requests for aproposed hydrogen control system as described in the SER. Amendment35 proposed the installation of a Containment Atmosphere Dilution(CAD) system. This system was to provide an onsite supply of nitrogenfor maintaining the containment atmosphere inert following thepostulated LOCA.
On June 13, 1974, the licensee informed NRC that work on the CADsyste'n was suspended pending revision of Regulatory Guide (RG) 1.7," Control of Combustible Gas Concentrations in Containment Followinga Loss-of-Coolant Accident."
Revision 1 to Regulatory Guide 1.7 was issued for comment in September1976; 10 CFR 50.44 was issued for comment as a proposed rule onOctober 21, 1976.
i On November 27, 1978, 10 CFR 50.44 became effective as a final ruleI and specified requirements applicable to those facilities, like the
Pilgrim facility, for which only a purging system was considered
necessary. This rule required that suitable redundancy be providedto assure that the safety function can be accomplished, assuming asingle failure, in accordance with Criterion 41 of Appendix A to 10CFR 50.
)On March 14, 1979, NRC requested.the licensee to submit, within 60days, a schedule for the installation of a CAD system, noting thatRevision l'to RG 1.7 had been issued for comment in September 1976and that Revision 2 to RG 1.7 was issued in final status in November1978. Furthermore, a detailed description of any design changesmade to the original FSAR submittal as a result of revised regulatoryguidance was also requested.
On June 6, 1979, the licensee informed NRC that "Our current plans |
do not call for the installation of a CAD System. We intend toretain the present inerted containment atmosphere requirements, andwe are evaluating a system that incorporates hydrogen recombinationcapability...Because of the extensive design changes resulting fromrevised regulatory guidance, we are unable to commit at this time toany detailed design change. We will submit a summary description ofour proposed system and our proposed schedule of implementation bySeptember 15, 1979."
Subsequently, on October 19, 1979 in BECo letter #79-207, the licenseeinformed NRC that they had evaluated the station design with respectto 10 CFR 50.44 and that " Based upon our analysis, we comply with 10CFR 50.44 with existing equipment." The licensee also requesteddeletion of FSAR Amendment 35 from their docket in that correspondence.The existing equipment included the nitrogen purge system which is apart of the containment atmosphere control system used for normaloperation to maintain an inerted containment and for post-LOCAcombustible gas control,
b. Design Evaluation / Modification Activities
On October 30, 1979, NRC requested that the liceasee submit, within60 days, analysis showing compliance with 10 CFR 50.44 as well asproposed Technical Specifications for post-LOCA hydrogen control.The licensee did not respond to the NRC request at that time; however,an analysis was internally documented by the licensee on March 28,1980, and was submitted as enclosure (A) to BECo letter #81-127 toNRC on June 15, 1981. The introduction to this analysis states inpart, "This analysis is the basis for the conclusion (in the October19, 1979 letter to NRC) that Pilgrim meets 10 CFR 50.44 with existingequipment. Subsequently, it was found that one of the assumptions(in the October 19, 1979 letter to NRC) was incorrect. It wasassumed that local operator action could be used for satisfyingsingle failure and loss of power criteria. A recent Reactor Buildinghabitability study, a result of the TMI Lessons Learned implementa-tion efforts, has demonstrated that the Reactor Building .,ay be
.._.
...
5
;
inaccessible after an accident...Because timely operator access forlocal action cannot be guaranteed, all 10 CFR 50.44 requirements arenot met with the existing equipment. Modifications are in progressto upgrade the system so that Pilgrim will comply with 10 CFR 50.44.These modifications are being implemented as quickly as possible."This failure to properly incorporate the requirements of 10 CFR50.44, although documented by the licensee in March 1980, was notreported to NRC until several months later.
Notwithstanding the licensee's statement o' October 19, 1979 and,
|- analysis documented in March, 198I., Amendment 35 had consideredpost-accident radiation levels it M e the reactor building anddocumented that reactor building accessibility was limited due tohigh dose levels. Consequently, remote operation from the controlroom was included in the design criteria for the proposed systemdesign. No reference was made to these prior studies and assumptionsin the aforementioned licensee's statement in 1979 or analysis ofMarch, 1980.
The modifications of the nitrogen purge system to preclude thenecessity for operator access to valves inside the reactor building;
are described in Plant Design Change Request (PDCR) 80-03, approvedfor implementation by the Onsite Review Committee (ORC) on March 21,1980 and PDCR 80-21, approved for implementation by the CRC on March19, 1980. The inspector reviewed the narrative portions of these
,
PDCRs and determined that the purpose and intent of the modifications; were adequately described. However, P&ID 6498-M-227 which was used: as a reference in preparation of the PDCR's had not been updated to
reflect changes made in system valve alignments by procedural changesin April,1979. The implications of this error will be explained in,
| subsequent paragraphs. The inspector noted that the offsite approvalsheets for these modifications indicated that they were also reviewedduring a pre-ORC meeting on February 15, 1980. The work on thesemodifications was completed during the May 1980 refueling out;tge;
i however, at the June 18, 1981 meeting between NRC and licenseei managers, it was determined that this system may not have beenj capable of being placed into service without the operator entering; the reactor building to open certain valves.
The flow path for the post accident nitrogen purge system is throughtwo one-inch local hand operated nitrogen makeup block valves (onetorus, one drywell) in the original nitrogen makeup system. These
,
;_ block valves were changed from the " locked open" to " closed" positionas a result of Revision 11 to System Operating Procedure No. 2.2.70,Primary Containment Atmosphere System on April 13, 1979. The statedreason for this change was to allow compliance with other proceduresand with current revisions of P& ids. This apparently facilitated
i routine operations, but no apparent consideration was given at the; time to the impact of this change on the capability to use this'
function in a post-LOCA situation. Emergency Procedure 5.4.6,Post-Accident Venting, was not revised to indicate that operators
!
|
i
_
- _ _ _ _ _ _ _ _ _ _ _ - _ _
-..
6
would be required to enter the reactor building in order to'open theclosed valves. fu'rthermore, P&ID 6498-M-227 was not revised to showthat the position of.the two inch local hand operated nitrogen blockvalves was changed from " locked open" to " closed." This failure tochange the P&ID appears to have had an adverse impact on futurerevisions which related to this system.
When Revision 11 to Procedure No. 2.2.70 was issued in April of1979, it in effect prohibited remote operation of the containmentatmosphere control system to allow post LOCA nitrogen purging of thecontainment. This revision, which required the makeup block valvesto be in the closed position, remained in effect until a TemporaryProcedure Change was issued on June 5,1981. Revision 17 to theprocedure, which incorporated the temporary change, was issued onJune 10, 1981. The failure to recognize the significance of Revision11 contributed to the following related discrepancies.
P&ID 6498-M-227 w:s not changed to reflect the altered position--
of the block valves. This appears to have resulted in inaccuraciesin the preparation of PDCR's 80-03 and 80-21.
-- Emergency Procedure 5.4.6, " Post Accident Venting" was notchanged to reflect the need for local operator action to openthe block valves in the event post accident containment ventingwas required.
-- The ORC reviews of PDCR's 80-03 and 80-21 held in March of 1980were improper since the reviews failed to take into account theimplications of Revision 11 to Procedure 2.2.70 which placedthe block valves in the closed position.
Additionally, the redundant post accident nitrogen purge systemrequired by 10 CFR 50.44 was totally disabled between July 23-24,1980 when the post accident nitrogen purge branch supply lines werecut and capped under an unauthorized and unapproved Field RevisionNotice Number 80-21-20. This maintenance quality assurance problemwas discussed in Inspection Report Number 50-293/81-13, and broughtto the licensee's attention in a Notice of Violation dated August11, 1981.
Other significant procedural inadequacies which were subsequentlyidentified and corrected include:
-- The design requirements of PDCR 80-03 stated that containmentpressure was to be limited to 50% of design (56 PSIG). WhenPDCR's 80-03 and 80-21 were implemented in May, 1980, thisrequirement was not reflected in emergency procedure 5.4.6,Post Accident Venting, which permitted containment pressures upto 40 psig. This discrepancy was not corrected until July 7,1981 when revision 9 to Emergency Procedure 5.4.6 was issuedwhich limited containment pressure to a range of 23 to 28 PSIGduring post accident conditions.
-- When the sixteen solenoid valves were installed through theimplementation of PDCR's 80-03 and 80-21, Procedure 8.7.1.5,Local Leak Rate Testing, was not revised to reflect the installationof these valves. This error was identified by the NRC ResidentInspector and is discussed in Inspection Report No. 81-12.Revision 10 to Procedure 8.7.1.3, issued by the licensee onAugust 14, 1981 corrected this discrepancy.
NRC inspection 79-09 conducted in May of 1979 reviewed in part,system valve lineup and locked valve status. During this inspection,it was noted that discrepancies existed with regard to valve statusin the Feedwater, High Pressure Coolant Injection and AtmosphereControl Systems. In regard to the latter system, the inspectornoted that the nitrogen addition block valves were in the lockedclosed position instead of the locked open position as specified bythe locked valve list. In two instances, licensee surveillances hadindicated that these valves were locked open when in fact they werelocked closed, one of them being red tagged closed. This resultedin issuance of an item of noncompliance. Two other items of noncompli-ance were issued as a result of valve status discrepancies identifiedin other systems. On October 2, 1979, the licensee responded to theInspection 79-09 findings and stated that for all safety' systems,the required valve positions (Appendix A) of appropriate 2.2 procedureshad been reviewed against the P&ID's, and that compliance had beenachieved with regard to agreement between 2.2 series procedures andrelated P&ID's. Apparently, the discrepancy with regard to thenitrogen block valve position between P&ID 6498-M-227 and Procedure2.2.70 already noted above, was not discovered by the licenseeduring this review.
c. Findings
(1) The licensee failed to provide a combustible gas control systemwhich met the requirements of 10 CFR 50.44. This violationresulted in plant operation without a system meeting the requirementsof the regulation from November 27, 1978 to June 5,1981. Thisis an item of noncompliance (81-18-01).
(2) The licensee failed to conduct an adequate safety analysis ofplant systems and procedures in April, 1979 as required by theprovision of 10 CFR 50.59. This resulted in the revision ofone procedure to close two manually operated nitrogen supplyvalves located inside the Reactor Building.
Post LOCA combustible gas control could not be assured sinceoperator access to the Reactor Building to open these valveshad not been evaluated. This is an item of noncompliance(81-18-02).
- ____ __ _ _ _-___- - _ _ -
_
...
8
.(3) When the licensee revised the operating procedure in April,1979 to close the two manually-operated nitrogen supply valves,he did not ensure that this change was reflected in the systemdrawing and an affected emergency procedure. This is an itemof noncompliance (81-18-03).
(4) Discovery during plant life of conditions not specifically-considered in the safety analysis report or technical specif-ications that require remedial action or corrective measures-toprevent the existence lor development of an unsafe condition, isa reportable occurrence requiring prompt notification of theNRC in accordance with Technical Specification 6.9.B.1.(i).Failure to report the discovery in March,1980, that the installednitrogen purge system did not meet the requirements of.10 CFR50.44, and that remedial action was required, is an. item ofnoncompliance (81-18-04).
(5) Licensees are expected to provide the NRC with information that-is accurate and complete. The Boston Edison Company letter ofOctober 19, 1979 to the NRC erroneously stated that compliancewith 10 CFR 50.44 was met with existing equipmer.t. This is-anitem of noncompliance (81-18-05).
3. Exit Interview
The resident inspector discussed inspection _ findings with plant managersas indicated below:
Date Item Person Contacted
July 7, 1981 Failure to Comply with 10 CFR R. D. Machon50.44
July 9, 1981 Failure to make required report C. Mathis
September 30, Failure to Comply with 10 CFR R. D. Machon1981 50, Appendix B, Criteria III
![0710 Eco (CFA520)[1]](https://static.documents.pub/doc/80x56/55cf8fce550346703ba00970/0710-eco-cfa5201.jpg)
