SNRC-1797, 10CFR 50.59 Annual Rept of Facility Changes,Procedure Changes,Tests & Experiments for Jan-Dec 1990

10CFR 50.59 Annual Rept of Facility Changes,Procedure Changes,Tests & Experiments for Jan-Dec 1990

ML20084U033
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	12/31/1990
From:	Leonard J LONG ISLAND LIGHTING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
SNRC-1797, NUDOCS 9103290067
Download: ML20084U033 (21)

Text

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[,#{g LONG ISLAND LIGHTING COMPANY

. , _ SHOREHAM NUCLEAR POWER STATION P C. BOX e18. NonTH COUNTRY ftO AD

• WADING RIVER N Y. I 1792 JOHN D LEONAftD, A vet PntSIDENT. OFFC( Of CORPORAtt $t8NCES AND VCt PRt8 DENT OFFCt O, NUCLt AR

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MAR 2 51991 U. S. Nuc1 car Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Submittal of the 10CFR50.59 Report for the Period January 1, 1990 Through December 31, 1990 Shoreham Nuclear Power Station - Unit 1 Dockej No. 50-322 Gentlement In accordance with the requirements of 10CFR50.59, this letter transmits Enclosure A, Shoreham Nucicar Power Station 10CFR50.59 Report for the period from January.1,_1990 through December 31, 1990. Title 10 CFR Section 50.59 requires that this report list those changes, tests and experiments which did not, by safety evaluation, involve an.unreviewed safety question and were completed during the reporting period.

Please be advised that 10CFR50.59 safety evaluations performed in support of Shorchem's System Layup Implementing Packages (SLIPS)~

program-are not included in the enclosed report. The purpose of the lay-up program is to protect from gross deterioration those systemo not required to meet Technical Specification _ requirements or-necessary to support minimal plant functions. Station Procedure (SP) S2.001.01, "SNPS Layup Program Description," which has been_ reviewed by NRC Staff and Region I personnel, provides i' details on the preparation, contents and approvals of system-layup implementation packages (SLIP), 10CFR50.59 safety evaluations associated with the SLIP program have been routinely reviewed by'NRC Staff and Region I personnel and these reviews have been documented in various Inspection Reports.(e.g., NRC f i Region I Inspection Report 50-322/90-01). /

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SNRC-1797

.Page 2 1 The format of the enclosed report is as follows: l SM/SPCN/SEPS No. - In the report, 10CFR50.59 items completed during the reporting period are listed by their Station Modification-(SM) number, Station Procedure Change Notice (SPCN) number, or Safety Evaluation Post Settlement (SEPS) ,

number. For convenience of reference, these are each listed separately in ascending order. Associated documents, i.e.,

Design Output-Packages (DOPs), are also given where applicable.

Description of Change - A brief description of the change, test or experiment addressed by the chango document.

Summary -: The safety evaluation determination that the change, test or experiment does not involve an unroviewed safety question pursuant to the three criteria of 10CFR50. 59 (a) (2) . ,

Should you require any additional information concerning this +

i submittal,.please-do not hesitate to contact this office. .

~Very truly yoursy

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(f th J I jhIlQ f/

fJoh* D. conar'd, Jr. '

/ Vic President, Office f Corporate Services

[L and Vice President, Off_ ice of Nuclear N'- ,-

Enclosure cc: S. Brown T. T. Martin B. Norris l

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SNRC-1797 ENCLOSURE A 10 CFR $50.59 Report Period From January 1, 1990 Through December 31, 1990 Shoreham Nuclear Power-Station Docket No. 50-322

,e LCRC-1797 Enclosure A Page 1 of 21 i SM 87-028 (DOP 87-044)

Description of Change This modification changed the Offsite Dose Calculation Manual (ODCM) software routine 1 in the Radiation Monitoring System (RMS) computer (ID11) which perform the offsite dose calculations. The change allows for accurate calculations when monitors measure i less than 20 cpm at RE-42, RE*21, and RE*22. '

l Summary I, ho. The' probability of occurrence or the consequences of an )

accident c*r malfunction of equipment important to safety <

previously_ evaluated in the Safety Analysis Report has not increased because the change only affects the method of calculation of offsite doses due to gaseous effluent releases from the plant. The modification does not affect the operation of the monitors or any safety function or safety related system.

II. No. A possibility for an accident or malfunction of a different type than any evaluated in the Safety Analysis Report has not been created. This modification does not affect the operation of any safety related system, it merely assures that the monitor's data are used properly.

III. No. The margin of safety as defined in the basis of any Technical Specification is not affected because the detectors only monitor plant effluents and do not affect any ,

plant limiting design conditions.

SNRC-1797 Enclosure A

, Page 2 of 21 SPCN 90-0250 ,

SPCN 90-0553 5FCN 90-0714 D_eg ription of Change SPCN 90-0250 revised Station Procedure (SP) 12.006.01, " Station Procedures - Preparation, Review, Approval, Change Review and Cancellation," by authorizing the appropriate Division Managers to assume approval authority for his/ncr designated procedures. ,

SPCN 90-0553 and 90-0714 revised Station Procedure (SP) 12.001.01, "Index and Organization of Station Operations Manual,"

by changing "P" approval 3evels to "D" approval levels, as appropriate. These revisions were made to achieve consistency with Technical Specification Section-6.8, " Procedures and Programs," of Shoreham's operating license NPP-82, as amended.

Summary I. No. These changes were administrative in nature and do not  ;

affect plant opecations: Changing the approval authority for procedures has no impact on accidents evaluated in the safety analysis report, nor doen it create the possibility for a new accident. Each Division Manager responsible for a given procedure has the qualifications, experience and familiarity-necessary to ensure that designated procedures approved by him/her comply.with regulatory requirements.

Other elements of procedure review, such as ROC review, are not affected by these changes.

II. No. See I above.

III. No. See I above.

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SURC-1797 Enclosure A

. Page 3 of 21 SPCN 90-0510 Description of Change SPCN 90-0510 created a Temporary Procedure (TP) 25.136.01, " Site Characterization of the RX Vessel." The purpose of this temporary proceduro is to coordinate the efforts of H.P.

Maintenance, Reactor Engineering, and Operations Sections in

, obtaining exposure data within the reactor (RX) vessel and its internals and provide guidance to allow site characterization of the reactor vessel.

Summary I. No. The defueled plant condition with fuel stored in the I Spent Fuel Storage Pool with the fuel pool gates installed eliminates safety concerns with respect to water level in the reactor vessel, movement of control rods in the reactor vessel, and the bypassing of level and control rod movement interlocks required for safety when fuel is in the reactor vessel.

II. No. See I above.

III. No. See I above.  ;

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SN RC-1 *. ,7 Enclosure S.

. Page 4 c; 21 SPCN 90-0945 Description of Change SPCN 90-0945 revised Station Procedure (SP) 12.014.05, " Required Reading List," by removing control of the required reading list from the Training Administrative Supervisor and placing the control with individual-Sections.

Summary I. No. The change places the control of distribution of the required reading list from the Training Department to appropriate Plant Staff Section lleads, tne change is administrative in nature and unrelated to any accident analysis and does not affect the function or operation of any plant system or equipment.

II. No. See I above.

III. No. See I above.

s SNRC-1797 Enclosure A Page 5 of 21 SPCN 90-1074 SPCN 90-1083 Description of Change SPCN 90-1074 and SPCN 90-1083 revised Station Procedures (SP) 27.421.01, "RBSVS and CRAC Water Chiller Load Capacity Verification Test," and (SP) 23.421.01, "RDSVS/CRAC Chilled Water," to test the chillers under as realistic conditions as possible and to set chiller maximum load consistent with expected maximum loads in Shoreham's defueled condition, respectively.

The load capacity of the M50 chillers was modified to reflect actual heat loads which exist with Shoreham in a defueleo/ fuel in the Spent Fuel Storage Pool condition.

Summary I. No. There is no increase in the probability of occurrence or the consequences of an accident or malfunction of I equipment important to safety previously evaluated in the safety analysis report. The probability of a chiller malfunction is actually reduced.

II. No. No physical alterations of plant configuration w<

ma02. The modi fication to load capacity enhances the performance of the chillers.

III, No. See I and II abrve.

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SNRC-1797 Enclosure A Page 6 of 21 SPCN 90-1172 SPCN 90-1173 SPCN 90-1174 SPCN 90-1177 SPCN 90-1178 Description of Change SPCN 90-1172 revised Station Procedure (SP) 61.010.01, " Access Controls."

SPCN r. 1173 revised Station Procedure (SP) 62.004.01, " Station Personnel Monitoring Program."

SPCN 90-1174 revi. sed Station Procedure (SP) 62.004.03, "Dioassay Program."

SPCN 90 ~177 revise 1 Station Procedure (SP) 61.010.03, " Health Physics Postinge and Sign'."

FDCN 90-1178 revised Station Procedure (SP) 61.012.01, " Personnel e Lir.its and Guides."

tion surveys generated in support of Shoreham's Site a s a. 'erization Program have revealed near background s ments throughout the accessible areas of the plant. To aLo the number of personnel that a.c required to wear des'. cry (TLD/DRD) inside the RCA, the above cited procedures we revised to require dosimetry (TLD + DRD) in only those areas rey iring radiation work permits (RWPs). The requirements for access to various areas have been revised to reflect these changes.

Summary I. - No. The changes to the procedures cited above are unrelated to any accident analysis and do not affect the function or operation of any plant system or equipment.

II. No. See I above.

III. No. See I above.

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SNRC-1797 Enclosure A-

, Page 7 of 21 SPCN 90-1376 Description of Change SPCN 90-1176 revised Station Procedure (SP) 12.014.03, " General Employee Training," to eliminate Visitor Training - Level II and change the requirement for GET II training to be consistent with changes made by SPCN 90-1172 and other associated SPCNs.

Summary I. No.- The changes to the procedure are unrelated to any accident analysis

• d do not affect the function or operation of any plant system or equipment.

II. No. See I above.

III. No. See I above.

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SNRC-1797 Ei. closure A Page 8 of 21 SPCN 90-1406 SPCN 90-1422 SPCN 90-1423 SPCN 90-1424 Description of Change As described in SNRC-1744 dated August 10, 1990, the Long Island Lighting Company (LILCO) completed a revision to the Shoreham Environmental Qualification (EQ) Program in accordance with 10CFR50.59. This EQ_ Program revision permitted (1) the deferral of EQ related preventive maintenance tasks and (2) the use of parts not qualified in accordance with 10CFR50.49 in applications where qualified parts would be used to withstand the harsh environmental conditions which could exist at an operating nuclear power plant. The technical justification for the EQ Program revision and safety evaluation were provided in SNRC-1744.

To attain consistency with the reviced EQ Program, the following Station Procedures were revised:

1) SPCN 90-1406 and 90-1423 revised Station Procedure (SP) 12.013.01, " Maintenance Work Requests."

2) SPCN 90-1422 revised Station Procedure (SP) 12.015.01,

" Preventive Maintenance Program."

3) SPCN 90-1424 revised Station Procedure (SP) 12.020.01,

" Environmental Qualification Program."

The revisions addressed the deferral of EQ re ated l maintenance and the use of parts not qualified in accordance with 10CFR50.49-and the record keeping function to track-all EQ requirements and the replacement of EQ equipment and/or components.

Summary The changes to Station Procedures cited above are within the scope of the safety evaluation provided in SNRC-1744.

I. No. Most of the accidents previously evaluated in the safety analysis report cannot occur with the reactor defueled and those that can 1) do not require equipment qualified in accordance with 10 CFR 50.49 to mitigate their consequences, and 2) have consequences that are well within the limitations of 10CFR100.

SNRC-1797 Enclosure A Page 9 of 21 SPCN 90-1406 SPCN 90-1422 SPCN 90-1423 SPCN 90-1424 (continued)

II. No. A review of USAR Chapter 15, a consideration of the defueled condition, and an assessment of the impact of the EQ Program modification does not reveal any conditions or circumstances that could result in an accident different than these already described in the USAR. There is no need for the EQ equipment while Shoreham is in the defueled, non-operating condition.

III. No. See I & II above, i

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SNRC-1797 Enclosure A ,

. Page 10 of 21 SPCN 90-1754 Description of Change SPCN 90-1754 created a new Station Procedure (SP) S2.001.02, "SNPS Decontamination Program," to provide the methods and administrative controls to decontaminate plant systems, equipment and-structures. The contamination level of Shoreham's structures, systems and components have been evaluated under the SNPS Site CI'aracterization Program. Those structures, systems, and componeats found with contamination levels above the acceptance c riteria of NRC Regulatory Guide 1.86 are being (will be) decontaminated in accordance with the methods and-controls of this procedure. Addendum 1 to the Site Characterization Program-will be used to control the radiological surveys as close as practical to the Final Termination Survey Plan as outlined in the Decommissioning Plan submitted by LIPA on December 29, 1990.

Summary I. No. Program Implementation utilizes a non-destructive chemical decontamination and survey program. The programs will be applied to systems not required for safe shutdown or to mitigate the consegue aces of an accident in the defueled plant condition. Program implementation is independent and unrelated to applicable accidents evaluated in the USAR in the defueled plant condition. The plant will not be operating'during the performance of this procedure. The execution of this procedure is a conservative action taken-to reduce contamination in preparation for decommissioning.

II. No. See I above.

III. No. See I above.

l SNRC-1797 Enclosuro A Page 11 of 21 SEPS90-004 Description of Change This change involved the re-assignment of the Radiochemistry /

ALARA and Health Physics Section Head of the Nuclear Engineering Department to the position of Manager, Radiological Controla Division (see SNRC-1711, April 6, 1990 and SNRC-1717, April 27, 1990) and the consolidation of organizational units within the Nuclear Engineering Department (NED) to more effectively utilize NED personnel for activities that are appropriate to the Shoreham Nuclear Power Station (SNPS) defueled condition and the requirements of a full power operating license with the overall functional capability of the Department being retained.

Specifically, the following organizational changes were involved:

1. The consolidation of the Reliability /PRA Section with the Safety Analysis Section into a new section entitled Reliability / Safety Analysis Section.

2. The consolidation of the Radiochemistry /ALARA/ Health Physics Section with the Nuclear Fuel Engineering Section into a new section entitled Radiation Protection / Fuels Section.

The organizational. changes, described above, were reviewed and ,

safety evaluation was prepared pursuant to 10CFR50.59 (a) (1) within the broad premise that plant organi2ation and personnel constitute, in part, "a segmunt of the facility as described in the safety analysis."

Summary i

I. No. The changes are administrative in nature and are unrelated to any accident analysic and do not affect the function or operation of any plant system or equipment.

The probability of accidents previously evaluated is not affected by these organizational changes. In Shoreham's non-operating, defueled condition, only the fuel drop accident and the radwaste tank rupture accident apply.

The continued. technical capability of NED is maintained to comply with license requirements.

II. No. No physical alterations of plant configuration or l changes to setpoints or operating parameters were made.

L The changes only reflect administrative control changes of various functional activities.

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i. III. No. The changes are unrelated to any margin of safety.

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SNRC-1797 Enclosure A Page 12 of 21 SEPS90-005 Description of Change As described in SNRC-1743 dated August 10, '990 this change eliminated the positions of Outage and Modifications Division Manager and Outage Engineer. Based on Shoreher,'s non-operating, defueled condition, and in accordance with the Agreement between LILCO and the State of New York that LILCO vill not operate Shoreham, the positions of Outage and Modifications Division Manager and Outage Engineer are not required because there is no longer a-need for planning of scheduled plant outages. The positions of Modification Engineer and Planning and Scheduling Engineer continue to retain their responsitilities as described in the Safety Analysis Report (SAR), but now report to the Operations Division Manager.

These changes, as identified in SNRC-174:3, were reviewed and a safety evaluation was prepared pursuant to 10CFR50.59 (a) (i) within the broad premise that plant organization and personnel constitute, in part, "a segment of the facility as described in the safety analysis."

Summary I. No. The changes are administrative in nature and are unrelated to any accident analysis and do not affect the function or operation of any' plant system or equipment.

II. No. No physical alterations of plant configuration or ,

changes to setpoints or operating parameters were made. The changes only reflect administrative control changes for

-outage functional activities.

III. No. The changes are unrelated'to any margin of safety.

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SNRC-1797 Enclosure A Page 13 of 21 SEPS90-007 Description of Change This change revised the USAR qualification requirement that the Operations Division Manager shall participate in management activities of an operating nuclear plant for two (2) months operation above 20% power and a routine refueling outage (1-2 months) by making them conditional requirements (i.e., the word "shall" was changed to "should"). Shoreham Technical Specification section 6.3.1 and USAR section 3B-1.8 state that the unit staff shall meet or exceed the qualifications of ANSI N18.1-1971. ANSI N18.1-1971 does not require the Operations Division Manager to have participated in the activities as stated above (and as identified in USAR section 13.1.3.2). The change does not delete the qualifications, but does revise the wording to eliminate it as a requirement. This change was made to support the information provided to the NRC via SNRC-1756 dated September 20, 1990.

Summary I. No. The change is unrelated to any accident analysis and does not affect the function or operation of any plant system or equipment.

II. No. No physical alterations of plant configuration or changes to setpoints or operating parameters were made.

III. No. The change of wording from "shall" to "should" is in accordance with Technical Specification section 6.3.1 and USAR section 3D-1.8. The Operations Division Manager must still meet the qualification requirements of ANSI N18.1-1971. The change is unrelated to any margin of safety.

SNRC-1797 Enclosure A Page 14 of 21 SEPS90-008 SPCN 90-0177 Description of Change A change _(modification) was made to RWCU pump G3J-P-019B in order to provide cooling to the pump using RWCU process water in lieu of RBCLCW. Additionally, to support this process, a chango (modification) was made to the non-safety related portion of the RBCLCW system which feeds the RWCU pump coolers. This overall change is temporary for plant layup purposes. SPCN 90-0177 revised Station Procedure (SP) 23.709.01, " Reactor Water Cleanup Syctem," to allow pump G33-P-019B to be operated with alternate cooling (RWCU) water.

The Shoreham plant is currently defueled with the reactor vessel being maintained at atmospheric conditions with a full water inventory.- Technical Specificati;n 3/4.4.4 requires that reactor water purity standards be maintained at all times. The RWCU system performs this task. However, the RBCLCW system has been laid up, and is therefore unavailable to provide cooling to the four cooling loads which are mounted on, and are a part of, the RWCU pump skid. In order to provide an alternate means of cooling and to allow the RWCU pump to operate, the subject modification was engineered. The cooling water is furnished by taking RWCU. water _from a vent in the pump's discharge, running it through the four cooling loads on the base of the pump, and returning it to the system via a test connection in the pump suction. RBCLCW is positively isolated from the modification by locking closed the isolation valves to and from the RUCU pump skid. Alternate cooling water is connected to the skid via inlet and outlet drain connections already provided, but which normally would be plugged. Cooling of two-loads, the bearing cooler and seal. cooler, is necessary for maintaining the pump operable.

(The-other two cooling loads, the pump cover and pedestal, are normally needed only when the reactor is operating.)

This modification allows _the RWCU system _to be used both for purification-of reactor vessel water as well as for water sampling to monitor that purity while the plant is defueled. The intent of this modification is that it only be employed while the reactor is at ambient conditions, and the levels of activity _in the coolant remain low.

The RWCU system is designed to process a fraction of reactor coolant flow during normal operation to remove solid and dissolved impurities in order to maintain water purity in accordance with Technical Specification 3/4.4.4. A small portion of the system forms part of the Reactor Coolant Pressure Boundary (RCPB). The cleanup portion of the system, that portion in which the modified pump is located, is not part of the RCPB, per USAR Section 5.1. During normal operation, the RCPB portion c f the

SNRC-1797 Enclosure A Page 15 of 21 SEPS90-008 SPCN 90-0177 (continued) system isolates in response to a high differential signal from the RWCU leak detection system, as well as in response to the-other signals described in the USAR. All of these signals are unaffected by this modification.

Summary I. No. This change (modification) does not affect any of the systems associated with previously evaluated accidents such that the probability of occurrence of one of those accidents, or the consequences of one of those accidents will be increased.

II. No. This modification is only used while the plant is in the defueled condition. Therefore, the pressures, temperatures and radiological considerations normally associated with an operating nuclear power plant are absent, and are of no concern for this modification. Also the cleanup portion of the-RWCU system, in which the subject pump is located, is not part of the RCPB. The use of process fluid for cooling outside the normal reactor coolant boundary, running it through a non-safety related cooling load, and returning it to the coolant loop, is of concern because water which would normally be considered reactor coolant is being removed from the piping designed to contain it at-operating pressures and temperatures, and then being returned to the coolant loop. Iloweve r , if a leak or rupture in the piping:should occur during the defueled condition, it would not pose a safety concern because the levels of

. activity in the process fluid being used are low. Because the plant is defueled, none of the added piping, or the existing low pressure cooling circuits on the pump baco, will be subject to the full reactor operating pressure. In addition, the 150 psig design pressure piping and cooling circuits on the pump base are protected from the maximum expected differential pressure produced by the pump through the use of a restriction orifice and a new relief valve, both conservatively designed to handle maximum anticipated conditions. Fir ally, the use of a Maintenance Work Request (MWR) requiring uismantling of the modification and

-decontamination of the affected RBCLW components coupled with the LCO attached to the RBCLCW system, will ensure that the original design configuration will be restored prior to any plant startup or fuel Joad activity.

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SIIRC-1797

' Enclosure A Page 16 of 21 SEPS90-008 SPCN 90-0177 (continued)

III. No. The margin of safety as defined in the bases of any Technical Specification will not be raduced as a result of providing alternate cooling to an RWCU pump while the plant is defueled. Thic is so because examination of the bases of Technical Specification sections 3/4.4.4 and 3/4.4.5, which discusses water chemistry, which is maintained during layup by the use of RWCU to remove particulates and dissolved impurities via the filter domineralizers, shows that providing alternate. cooling to an RWCU pump while the plant is defueled does not affect these bases.

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SNRC-1797 Enclosure A Page 17 of 21 SEPS90-009 Description of Change The operating methodology for the RBNVS was revised from a " flow thru" system.to a " recirculation" system while SNPS is in a defueled mode. This mode of operation of the Reactor Building HVAC system allowed:

1. The supply portion of the RBNVS to operate in a recirculating mode.

2. The RBSVS cooling portion only to operate.

3. The exhaust portion of the RBNVS to operate partially.

The change was done in order to maintain a 35% relative humidity and'a 1/4" - 1" w.g. negative pressure within the Reactor Building during the defueled mode of operation only.

Simultaneous operation of RSNVS and portions of RBSVS necessite.tes.that certain-leads to dampers and cooling coil control valves be lifted in order to make.the control of these dampers independent of the parent system. In order to operate the RDNVS supply system in a recirculating mode to provide good  !

mixing of the air and to provide heating during the winter, th 3 Reactor Building isolation valves in the RBNVS outside air intake (1T4G*AOV-35A/B) will be closed while the access doors are open.

To maintain-a negative pressure of.1/4".- 1" w.g. within the Reactor Building requires.that the isolation valves in the RBNVS exhaust (lT4 6

• A0V-3 7 A/B ) be lined up such.that one valve is fully open while the second valve in series is cracked open sufficiently to utilize the exhaust fan " head" to create a negative 1/4"'-~1" w.g. pressure. The new operating mode with an

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exhaust air flow rate of approximately 1160 scfm compared to the.

normal flow rate of 90,000 scfm will. affect the operation of i 1D11-PNL-029. The isokinetic nozzle in the duct will no longer-function as intended since the withdrawal rate of the isokinetic nozzle will be too rapid for the reduced exhaust flow rate.

However, virtually all iodines have decayed away, and,-with the fuel'in the pool, particulates should be retained in the fuel-pool. Furthermore, the. monitors in this panel do not serve as-effluent monitors or fulfill a safety related function. They are

" intended to detect airborne radioactivity in-the Reactor

= Building . -However,=-there are no detectable gaseous sources at SNPS, either present or. anticipated, and the Semi-Annual Effluent-Release Report indicates there were no releases from the various building exhaust systems. Any releases from the fuel in the fuel pool:would be detected by the safety related refueling level exhaust monitors 1D11-FE-17A/B and 1Dll-RE-041 in the station vent. 11Dll-RE-029 noble gas monitor would also detect releaces, since noble gas monitoring is not affected by deviations from

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SNRC-1797 Enclosure A Page 18-of 21 SEPS--90-009 (continued)-

isokinetic sampling, and the noble gas Kr-85 would be the predominant isotope in airborne releases from the fuel.

Therefore, the loss of the isokinetic function of the 1Dll-PNL-029 nozzle has no safety significance in the current

.defueled condition. To provide a 35% relative humidity requires that the air circulating within the building be cooled to remove the moisture.- The RBSVS cooling coils are therefore required to Loperate. Modifications to the electrical and instrument air system were necessary to operate the RBNVS and RBSVS as stated above.- These changes are fully' reversible. The design bases of the RBNVS was changed by this modification.

Summary I. N o '. The change in operation of the RBNVS is only acceptable for plant operation in the defueled modo, and must be raversed for any other mode of operation. Therefore, the modification does not affect the functionality of any system intendedqfor safe shutdown of the reactor, or required to mitigate'the consequences of an accident, nor does it entail any degradation of any system related-to the reactor coolant pressure boundary or primary containment. To operate in any other condition than "defueled" requires that the changes be reversed and hence the functionality of any system intended toimitigateithe effect of an accident will not be affected.

II. No. :See I above.

-III. No. -See'I above.

t SNRC-1797 Enclosure A Page 19 of 21 SEPS90-010 Description of Change As described in SNRC-1771 dated November 13, 1990 this change announced a number of Quality Assurance Department organizational changes at the Shoreham Nuclear Power Station. The changes, as identified.in SNRC-1771, were reviewed and a safety evaluation was prepared pursuant to 10CFR50.59 (a) (1) within the broad premise that the plant organization and personnel constitute, in part, "a segment of the facility as described in the safety analysis." -

These changes are as follows:

.The Nuclear. Quality Assurance Department (NOAD) has been integrated.with the LILCO Corporate Quality Assurance Organization. The Manager, NQAD will report to the Manager of_ Corporate Quality Assurance for policy matters, personnel administration, coordination of resources and budgetary control. However, the direct operational control and day to day reporting remains, as in the past, to the Vice President

- Office of Corporate Services and Vice President - Office of Nuclear. The Manager NOAD continues to have a direct line of communication to the President of the Company for matters concerning nuclea. =afety.

The Manager, NQAD will remain as Chairman of the Nuclear Review Board which.will continue to function. The NOAD will continue to be located at the SNPS site and has full responsibility for Nuclear QA functions.

-The' Safety Engineering and Reliability (SER) Division, since it is;no longer required at the Division level, has been removed'from the organizational structure. As appropriate this Division's responsibilities have been reassigned or determined to be no longer pertinent. The Reliability Engineering Section is the only. functional area to be discontinued. This secticn previously performed reliability studies of plant systems and components and was a non-regulatory required group which had the objectives of increasing plant operational reliability, availability and

. maintainability. Under Shoreham's existing non-operational and defueled condition, these objectives serve no useful purpose.

The Independent Safety Engineering Group (ISEG), which was previously part of the Safety Engineering and Reliability Division, will report directly to the Manager, NOAD, pending NRC approval of License Change Application #8 which requests permission to eliminate the ISEG.

j SNRC-1797

- Enclosure A I Page 20 of 21 SEPS90-010 (continued)

> Summary.

I. No. These changes are administrative in nature and are unrelated to any accident analysis and do not affect the function or operation of any system or component. No program' commitments are reduced. The change does not eliminate NOAD, but merely changes the organizational reporting lines.

II. No. No physical alteration of plant configuration or

- changes to setpoints or operating parameters occurred. The changes do not affect the function or operation of any plant system or equipment. The changes reflect administrative control changes of various functional activities.

III. No. _The' changes are unrelated to any margin of safety.

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SNRC-1797

+- Enclosure A i Page 21 of 21 SEPS90-012 Description of Change This change . (ECR H-00727) provided that normally manually closed '

valves 1G41-04V-0068, 1G41-04V-0069, and.lG41-06V-0070 in the Fuel Pool Cooling a..d Cleanup System would be locked closed. l This change was made in accordance with INPO recommendation (INPO  !

SEN-20,-9/25/87) to prevent inadvertent draining of the Spent Fuel Storage Pool.

I Summary I. No. This change was administrative in nature and does not affect plant operations. The change does not involve any change to the design basis of the plant or any modification to plant structures, systems, or components. The change is unrelated to Any. accident analysis and does not af fect the function or operation of any plant system or equipment.

II. No. See I above.

III. No. See I above.  !

l t

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