SNRC-1904, Forwards 10CFR50.59 Rept for Period from Jan-Dec 1991
| ML20141M126 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 03/30/1992 |
| From: | Leslie Hill LONG ISLAND POWER AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LSNRC-1904, NUDOCS 9204010252 | |
| Download: ML20141M126 (27) | |
Text
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Lo g Shoreham Nuclear Power Station
[,
Island P.O. Box 628 i 'p.
pota Nonh Country Road Wading River, N.Y.11792 Acamnty LSNRC-1904 W92
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U.
S.
Nuclear Regulatory Commission ATTN:
Document Control Desk I
Washington, D.C.
20555 Submittal of the 10CFR50.59 Report for the a
Period January 1, 1991 Through December 31, 1991 Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322 Gentlemen:
This letter transmits Enclosure A, Shoreham Nuclear Power Station 10CFR50.$9 Report for the period from January 1, 1991 through December 31, 1991.
During the period addressed by this report, Long Island Lighting Company (LCLCO) was owner and licensee of the Shoreham Station under license NPF-82, as amended.
Title 10 CFR Section 50.59 (b) (2) requires that this r port list those changes, tests and experiments which did not, by safety evaluation, involve an unreviewed safety question and were completed during the reporting period.
The activities addressed by the enclosed report (1) were permitted by the Shoreham License (NPF-82) and 10CFR50.59 and (2) do ret foreclose options or materially affect the cost of decommissioning.
The format of the enclosed report is as follows:
SPCN/SE No. - In the report, 10CFR50.59 items completed during the reporting period are listed by their Station Procedure Change Notice (SPCN) number, or Safety Evaluation (SE) number.
For convenience of reference, these are each listed separately in ascending order.
Description of Change - A brief description of the change, test, or exper. ment addressed by the change document.
Summary - The safety evaluation determination that the change, test or experiment does not involve an unreviewed safety question pursuant to the three criteria of d
I 10CFR5 0. 59 (a) ( 2).
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Martin B. Norris 1
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I LSNRC-1904 Enclosure A 10 CFR 950.59 Report Period From January 1, 1991 Through December 31, 1991 4
Shoreham Nuclear Power Station Docket No. 50-322 Q
4 LSNRC-19 0 4 Enclosure A Page 1 of 24 SPCN 93-0064 Description of Change SPCN 91-0064 revised Station Procedure (SP) 23.104.01,
" Condensate Demineralizers," by adding three (3) new appendices:
1)
Appendix 12.9
- System N52 Regenerated Resin Transfer From Resin Storage Tank To Vendor Supplied Ion Exchange Column.
Appendix 12.10 - Valve Line-Up Checklist For Posin Transfer Using P-149 To Vendor Supplied Ion Exchange Column.
Appers'
.2.11 - System N52 Regenerated Rosin Transfer From Rosin Storage Tank To Tho Pacific Nuclear Services Ion Exchange Column.
Addit.otaSly, a new NOTE 2 was added to Section 3.0, a new Section 5.3 was added, and Section 12.0 was revised to include the now appendices and address their use.
To support the chemical decontamination of plant systems, it was determined that it would be necessary to transfer regenerated resin from the Resin Storage Tank to a vendor-supplied Ion Exchange Coluron in the Radwaste Truck Bay.
The addition of Appendices 12.9, 12.10, and 12.11 permits this resin transfer.
Summary I.
No.
The revision to SP 23.104.01 is unrelated to any accident analysis described in the Safety Analysis Report and does not affect the function or operation of any plant system or equipment.
Shoreham's plant condition is defueled, with all fuel stored in the Spent Fuel Storage Pool and fuel pool gates installed.
The System Layup Implementation Program is in effect for the protection of various plant systems.
This procedure revision meets the design basis of the External Regeneration System.
The new 8
Appendices will be used only as needed for resin transfer and wil.1 not ca:.se c permanent alteration in the design of any existing plant systems for normal operation.
During performance of these Appendices, resin transfer from the Resin Storage Tank to the Spent Resin Tank is prohibited.
In using this revised procedure, Shoreham will be transferring regenerated (i.e.,
clean) resin.
Analysis has revealed concentrations of Co60 and Mn54 of very low activity.
The dose rate of the resin is less than.5 mroms/hr. on contact and does not present. any radiological impact to the offsite general public.
Any leakage or resin spills are handled using existing Health Physien procedures.
II.
No.
See I Above.
LSNRC-1904 Encionure A Page 2 of 24 III. No.
All applicable technical specifications are adhered to when transferring resin.
The margin of safety as defined in the basis for any technical specification will not be reduced by performance of the added Appendices.
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1.SN RC-19 0 4 l
Enclosure A Page 3 of 24 SPCN 91-0078 Description of Change SPCN 91-0078 created a new Station Procedure (SP) S4.709.01, "RWCU System Dilute Chemical Decontamination Procedure," to provido station operating personnel and nystem engineers detailed instructions for decontaminating the Reactor Water Cleanup (RWCU) system.
Summary I.
No.
The RWCU system in an auxiliary system and not required to operate to mitigate or combat an accident, Application of dilute chemicals (CITROX Proccan) to decontaminate the RWCO system is independent and unrelated to applicable accidents evaluated in the Safety Analysis Report in the defueled plant condition.
The plant will not be operating during the performance of thin procedure.
The execution of this procedure is a conservative action taken to reduce contamination in preparation for decommissioning.
II.
No.
. lut dilute chemical decontamination application is carried out in compliance with existing regulations and approved administrative controls.
The application of an accepted chemical decontamination process, specimen sampling and evaluation and administrative controls collectively provide reasonable assurance that the possibility for an accident or malfunction of a different type than any previously evaluated is not created.
I I 7.. No.
Conformance to the applicable technical specifications and regulatory requirements during the application, handling and shipping of dilute chemicals ensures that the margin of safety as defined in the basis for any technical specification is maintained, s
6 3
LSNRC-1904 Enclosure A Page 4 of 24 l
SPCN 91-0164 r
Description of Change SPCN 91-0164 created a new Station Procedure (SP) S4.709.02, "RWCU System Pump Operation With The Reactor Vessel Drained", to provide guidelines for a flow path for the Reactor Water Cleanup (RNCU) system pumps and operation of the pumps with various flow paths to filter out any material that is in the system that can be removed by water circulation with the reactor vessel drained.
Summary I.
No.
Shoreham is in a defueled condition with the all fuel stored in the Spent Fuel Storage Pn,1.
The Reactor Water Cleanup system has been chemically macontaminated using the CITROX decontamination process.
The RWCU system is an auxiliary system and does not mitigate or reduce the consequences of any accident in the defueled plant condition.
The RWCU system pumps are not safety-related and the consequences of a malfunction of the RWCU system components has no impact on safety related equipment needed to mitigate or reduce the consequences of the credible accidents analyzed for the defueled plant condition.
II.
No.
The major difference in operating the RWCU pumps under this procedure is that water is contained within the system and not being taken from or returned to the Reactor Coolant system.
Running the RWCU pumps in a closed loop of the RWCU system, essentially an isolated system from the Reactor Coolant system, does not create the possibility for an accident or a malfunction of a different type than any previously analyzed.
III. No.
The margin of safety as defined in the basis for any technical specification will not be reduced by operating the RWCU system in the configuration addressed by this procedure.
LSNRC-1904 Enclosure A Page 5 of 24 SPCN 91-0724 Description of Change SPCN 91 -0724 created a new Station Procedure (SP) S4.203.01,
" Core Spray Dilute Chemical Decontamination Procedure", to provide station operating personnel and systen engineers detailed instructiono for chemical decontamination of the Core Spray (CS)
System.
Summary I.
No.
In Shoreham's non-operating, defueled condition, the Core Spray (CS) system is not required to operate to mitigate or combat an accident.
Application of dilute chemicals (CITROX Process) to decontaminate the CS system is independent and unrelated to applicable accidents evaluated in the Safety Analysis Report 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.
The dilute chemical decontaminat-ion application is carried out in compliance with existing regulations and approved administrative controls.
The application of an accepted chemical decentamination process, specimen sampling
^
and evaluation and administrative controls collectively provide reasonable assurance that the possibility for an accident or malfunction of a different type than any previously evaluated is not created.
III. No.
Conformance to the applicable technical specifications and regulatory requirements during the application, handling and shipping of dilute chemicals ensures that the margin of safety as defined in the basis for any technical specification is maintained.
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LSNRC-1904 Enclosure A 3
Page 5 of 24 SPCN 91-0724 Description of Change SPCN 91-0724 created a new Station Procedure (SP) S4.203.01,
" Core Spray Dilute Chemical Decontamination Procedure", to
/
provide station operating personnel and system engineers detailed instructions for chemical decontamination of the Core Spray (CS)
System.
Summary I.
No.
In Shoreham's non-operating, defueled condition, the Core Spray (CS) system is not required to operate to mitigate or combat an accident.
Application of dilute chemicals (CITROX Process) to decontaminate the CS system is independent and unrelated to applicable accidents evaluated in tb.
Safety Analysis Report 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.
The dilute chemical decontamination application is carried out in compliance with existing regulations and approved administrative controls.
The application of an accepted chemical decontamination process, specimen sampling and evaluation and administrative controls collectively provide reasonable assurance thst the possibility for an accident or malfunction of a different type than any previously evaluated is not created.
III. No.
Conformance to the applicable technical specificaticns and regulatory requirements during the application, handling and shiyging of dilute chemicals ensures that the margin of safety as defined in the basis for any technical specification is maintained.
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LSNRC-1904 Enclosure A Page 6 of 24 SPCN 91-0899 SPCN 91-0919 Description of Change SPCN 91-0899 and SPCN 91-0919 revised, respectively, Defueled Emergency Preparedness Implementing Procedures (DEPIPs) 5-8, "Communientions Operations cnd Tests", and 1-5,
" Notifications",
by removing references to the Emergency Notf.fication System (ENS).
The ENS was designed and installed as a dedicated communications line to the NRC in the unlikely event of an emergency at i
Shoreham.
No dialing was necessary to establish communications.
The NRC was financially-responsible for this system.
Due to the defueled status of Shoreham, and the cost savings afforded to the NPC, the NRC has decided 'co remove the ENS from Shoreham.
The removal of the ENS does not, however, alleviate Shoreham from the notification requirements of 10CFR50.72 or testing requirements of 10CFR50, Appendix E, Section IV.E.d.
Notification and testing requirements are now met using a commercial phone.
Summary I.
No, These revisions were administrative in nature and 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.
- ' ' - ~ ' - - ' - - ' ~ - -
LSNRC-1904 Enclosure A Page 7 of 24 SPCN 91-1191 Description of Change SPCN 91-1191 created a new Station Procedure (SP) S3.421.01,
" Protected System Lineup For M RBSVS and CRAC Chilled Water System", to provide guidelines for placing systems into i PROTECTED status per the Shoreham Nuclear Fower Station Layup Program.
The Procedure provides appendices for system draining and dryout, PF.OTECT]D valve / electrical / annunciator checklists,
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and other appendices as required for Shoreham*s System Layup i
Implementing Package implementation.
Summary I.
lo.
The implementation of this new Procedure S3.421.01 is l
consistent with the Possession Only License, and associated Technical Specifications, issued to Shoreham on June 14 1991.
The probability of occurrence or the consequences, of an accident or malfm etion of equipment important as evaluated in the Defueled Safety Analysis Reportto safety has not been increased because the Control Room AC (DSAR)
(CRAC) system remains unchanged in design and operating functions.
liowever, the system is reclassified to QA Category II; filter portion of the system is no longer required and one the of each of the redundant required.
fans and AC units is no longer environment for the operators.The AC system only functions to provide an OSIIA This requires the operation of only one RBSVS System)/CRAC chiller.(Reactor Building Standby Ventilation Automatic initiation systems and interlocks for.the habitability portion of the system are non-operable and the AC system is manually controlled from the Control Room.
Only one of the redundant trains is required.
There are two (2) water chillers in each redundant train of the M50 system.
af fects the layup of one of the chillersThis procedure only train.
(1M50*WC-003B) of the "B"
II.
No.
Only one chiller is now required in Shoreham's
- defueled, non-operating condition.
Chillers 3A, 4A and 4D remain in service to provide chilled water to the RBSVS and CRAC systems.
III. No.
See I and II above.
LSNRC-1904 Enclosure A Page 8 of 24 SPCN 91-1403 3PCN 91-1490 SPCN 91-1491 SPCN 91-1492 Description of Change SPCN 91-1483 deactivated Station Procedure (SP) 29.021.01, " Loss of Turbine Building Closed Loop Cooling Water (TBCLCW)."
SPCN 91-1490 deactivated Station Procedure (SP) 29.016.01, " Loss of Instrument Air-Emergency Procedure."
SPCN 91-1491 deactivated Station Procedure (SP) 29.023.02,
" Secondary Containment Control-Emergency Procedure."
SPCN 91-1492 deactivated Station Procedure (SP) 29.023.06,
" Radioactivity Release Control-Emergency Procedure."
In Shoreham's defueled plant condition with all fuel stored in the Spent Fuel Storage Pool, and the Possession Only License in effect, there is no need to maintain the Emergency Operating Procedures specified in Defueled Safety Analysis Report (DSAR)
Table 13.5.1-1, items B.3.g, B.3.1, B.3.j and B.3.k.
The emergency actions specified in these deactivated procedures are either no longer pertinent to Shoreham, or have been transferred to the abnormal performance section of the normal System Operating Procedure.
Summary I.
No.
In Shoreham's defueled condition, there are no active safety systems required to mitigate the consequences of an accident.
The fuel is stored in the Spent Puel Storage Pool which-provides a high degree of passive safety.
- TBCLCW, Plant Air, and Secondary Containment are not required to mitigate accidents.
Based on the Defueled Emergency Preparedness Implementing Procedures,
.69 highest Emergency Action Level that is required at Shoreham is an " ALERT".
The existing Alarm Response Procedures together with the Offsite Dose Calculation Manual and Process Control Program contain the necessary required actions in the event of a radioactive release.
Deactivating the above cited procedures will not increase the probability of occurrence or. consequences of an accident previously evaluated in the DSAP.
II.
No.
There is no possibility that an accident or malfunction of a different type than previously evaluated in tne DSAR will be created by deactivating the procedures.
III. No.
See I and II above, m._
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LSNRC-1904 Enclosure A Page 9 of 24 i
SPCN 91-1568 Description of change FPCN 91 1068 revised Defueled Emergency Preparednous Implomonting Procedure (Di. PIP) 1-9, Technical Support Contor (TSC)
Activation", by-ramoving references-to the MET Data Recorder.
The MET Data Fecorder contained signals from the 33 foot MET tovfor whic). una eliminated.
(Picase refer to Section F of the j
flemiannual Radioagtive Ef fluent holease Report submitted by LILCO lotter SNRC-1902 - dated Pt,bruary 27, 1992 for further information on-the-33 foot MET tower.)
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i Summary l
'I.__
No.
This revision is administrative in nature and conforms to the Technical Specifications issued with the Possession
~
Only License.
The revision-is unrelated to any accident analysis and does not affect the function or operation of any plant syetem or equipment, j
II.
No. - See I above.
III. No.
See I abovu.
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Enclosure A Page 10 of 24 GPCN 91-1776 Description of Change SPCN~91-1776 created a Temporary Procedure (TP) S4.122.01, "TBSW Special Test".
Tho-purpose of this temporary procedure is to este.blish a minimum (loss than 100 gpm) flow t1 rough the operating Turbinc Duilding Closed Loop Cooling Wa?.or (TBCLCH) hoot exc0 anger to determine the feasibility of installing and operating temporary serv $ce water pumps of significant lower
- capacity, in the current defueled-plant cendition, with minimal heat loads in the TBCLCW system, full design Llow through the TBCLCW heat exchangers.should not be required.
This temporary procedure provides direction for installing jumper hoses-around the operating TBCLCW heat exchar. gor service water outlet MOV
. (IP41-MOV-111A or MOV-111U).
With the jumpers installed, the outlet MOV will be cJosed thereby fting service water flow through: the heat exchanger to tb' cap tity of the jumper hoses.
Jice Water (TBSW) flow With the--minimum Turbine Building..
established, the TSCLCW temperature wl.11 bo monitored and recorded to determine the feasibility of operating in this mode.
IT-satisfactory operation of the TBCLCW con be indefinitely
- maintained with reduced TBSW flow, a determination will be made ao to whether or not to install smaller pumps to operate during routine conditions in lieu of running the Reactor Duildinc Service Water-(RUSW) pumps.
One PSSW pump will-remain functional to support operabili;y of a non-safety diesel generator during fuel movement, per technical specification requirements.
Summary I.
No, In Shoreham's defueled condition, there are'no active safety systems required to mitigate the consequencas of an tecident.
TDCLCW is not required to mitigato accidents defined by the Defueled Safety Analysis Report (DSAR).
i Performance of the TBSW. test, in accordanae with the
' temporary procedure, will not increase the probability of occurrence-or-consequences of an accident _previously evaluated in the DSAR.
= 11.
No.
The possibility for an accident or malfunction of a different type than any evaluated previously in the Defueled Safety Analysis Report is not created by performance of thic test.
III.-No.
See I and II abovc.
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Enclosure A Page 11 of 24 1
l DPC.N 91-18_8_1_
Description of Change l
lSICN 91-1881 created 6 Temporary Procedure (TP) 81.120.01,
" Administrative Procedure For Evaluating Initial B31 System Pipe a
Cut".
The purpose of this temporary procedure is to provide a means for Leng Island Power Authority (LIPA) to evaluate the adequacy of existing LILCO Radiological Controls Division procedures and their applicability to LIPA decommissioning work i
activities, and to.provido LIPA.a means of evaluating a major I
i aspect of the LEPA decommissioning physical work process.
The Reactor Water-Racirculation system (B31) includes those I
systems and components that.contain or transport fluida coming from or going to the rea; tor core.
In Shoreham's defueled condition, the fuel in not_in the core and the reactor is r
depressurized.
Therefore, the Reactor Water Recirculation system is not required.
This system in a non-operablo system and has been reclasvified "non-safety.related, GA Category II" by the Defueled Safaty Analysis Report (DSAR).
Removing a section of
-pipe from the D loop of this system will not affect any operational or' functional system of the plant.
j Summary I.
No.
The performance of this procedure will not affect existing eafety related components, systems or structures.
~
The Reactor Water Recirculation system, in accordance with the.DSAR, is classified non-nafety related, QA Category II, and not operable.
Removing _ a section of pipe will not increase the probability of1 occurrence or-consequences of an accident or increase the prok-bility of malfunction of-equipment important to safety.
11.
No.
Removal of a section of pipe from the system will not affect any other-safety related components, systems or structures.
The--system is drained and not operable, and classified non-safety related, QA Category II.
III. The margin of safety as defined in the basis for any defueled technical. specification is not reduced since the performance of this procedure has no effect on any safety
-relat3d component, system or structure.
The' performance of-r this procedure-will not cauce im of f site dose relense in excesb of that_ established in the USAR/DSAR.
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r Erclosure A Page 12 of 24
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DE 91-001 Description of-Change l
i This safety evaluation was performed to provide the basis for determining that the replacement of some copper-nickel (Cu-Ni) l spool pieces with Honel pieces in the P41 (Service Water) portion of-the M50 (RBSV and Control Room Chilled Water) system did not-involve an unreviewed safoty question.
The affected piping involves those spool pieces which are located immediate1'/
downstream of the temperature control valves at the outlet of the condensers.
r To prevent accidental shutdown of the water chillers due to low cooling water teeparature,-a mixing valve is provided to ensure that cooling water temperature entering the chillers is at least i
70*r.
Downstream of-thin mixing valve __and the acsociated
+
__ overboard temperature _ control valves, erosion has resulted in piping thinning.
The cause of the crosion has been determined to be excessive velocitics_immediately downstream of the temperature control _ valves resulting from valve throttling daring the winter and summer months.
This throttling, coupled _with entrained solids, has resulted in spool piece degradation.
1
- SummaIy, I.
No.- All safety system operating parameters remain unaffected.
Chiller availability is enhanced by this material improvement by providing granter assurance of system operability and integrity.
Seismic integrity is maintained for all components and piping associated with the
-M50 cbt11ers.
II.
No.
See I above.
III. No.
System design and operating conditions as set forth in the Safety Analysis Report are not altered by this replacement.
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LSNRC-1904 Enclosure A Page 13 of 24 4
I SE 91-004 Description of Change This safety evaluation was prepared to provide the basis for determining that the replacement of semo copper-nickel (Cu-Ni) spool pieces with Monal pieces in the P41 (Service Water) portion for the TDI Emergency Diesel Generators (EDGs) did not involve an unreviewed safety question.
The affected piping involves those
~
spool pieces which are located immediately downstream of the
. manual valves downstream of the TDI EDG jacket water heat exchangers.
During operation of the diesel engine, the air operated valve (1P41*A0V16) opens and permits service watet flow through the jacket water heat exchanger.
To regulate the engine's jacket water temperature, the manual valve downstream of the air operated _ valve is throttled.
Downstream of this manual valve, erosion has resulted in pipe thinning.
The ecuse of the erosion has been determined to be excessive velocities immedittely downstream of the manual valve resultir.g _ from valve throttling-during diesel engine operation.
This throttling, coupled with entrained solids, has resulted in spool piece degradation.
Summary i
I.
No.
All safety syscem operating parameters remain unaffected.
Emergency Diesel Generator availability is enhanced by this material improvement by providing greater assurance of system operability and integrity.
Seismic integrity is maintained for all_ components and piping
. associated with the TDI Emergency Diesel Generators.
l
- 11. - No.
See I above.
III. No.
System design and operating conditions as set-forth in 7
the Safety Analysis Report are not
'tered by this replacement.
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LSHRC-1904 Enclosure A Page 14 of 24 Sh 91-009 Description of Change.
This safety evaluation was performed to provide the basis for determining that reduction in the Radiological Environmental I
i.e.,
elimination of milk and ground Monitoring Program (REMP) water sampling, elimination of I-131 analysis, and elimination of l
the outer ring of direct radiation monitors, controlled by the l
Offsito Dose Calculation Manual (ODCM), did not involve an unroviewed safety question.
Regulatory Guide 4.1 indicates that if a facility can demonstrate z
that either-a particular set of isotopes and/or pathways.no
- longer exist, or, even if they do exist, there is no measured radiological impact toEthe public, then these isotopes / pathways
- need not be sampled.
l Since ShorehamElas_t generated power in-June 1987 (last 5.0%-power l
test) and last achieved stiticality in January 1989 (reactor operator training), - tho generation of fission
, activation and/or. transuranic - products was offectively terminated in June 190',_and radioisotopics generated prior to this date have undergone four years of decay.
Short-lived isotopee, such as I-131, have completely decayed away.
Based on this information, justification existed to reduce REMP below operational technical specification requirements consistent with Regu3atory Guide 4.1.
The scope reduction of REMP, reflecting the non-operating and defueled status of Shoreham, was identified to the NRC in Section H,
Miscellaneous Special Reports", of the Semiannual Radioactive
[
= Effluent Release Report forwarded by LILCO letter SNRC-1825 dated August 29,_1991.
Summary I.
No.
REMP' addresses after-the-fact releases and does not af fect-the pcobability of occurrences or the consequences of an. accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report.
REMP does not intaract with any safety-systems at Shoreham, and does not affect Shoreham's ability to remain in a safe shutdown condition with-all fuel stored in the Spent-Fuel
-Storage Pool.
11.
No.
A possibility for an accident or malfunction of a different type 1than-any evaluated in the Safety Analysir Report has not been created.
The revision to REMP does not affect the operation or function oi-any safety related systems.
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I LS!IllC-19 0 4 1:nclosure A Page 15 of 24 III. 11 0.
The margin of safety as defined in the basis of any technical specifiction is not affected because no safety cyctem J.s affecte(1 and the revision does not violato any of the bases in the Technical Specifications or the ODCM.
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a LSNRC-1904 Enclosure A Page 16 of 24 SM 91-010 Description of Change This safety evaluation addresses Revision 3 of the Shoreham Defueled Safety Analysis Foport (DSARI which deletes OA Category IIA end reclassifies OA Category IIA systems, structures and components as QA Category II.
The safety _ evaluation also addresses editorial changes which are updates, corrections and clarifications.
The safety evttluation was prepared to provide the basis for dctormining that redesignating QA Category IIA
-systems, structures, and components to QA Category II did not involve an unreviewed safety question.
g Revision 3 of the DSAR was forwarded to the NRC by LILCO letter SNRC-1820 dated August 26, 1991.
Summary I.-
No.
Relief from the stringent tracking requirement for spare parts and modifications of non-safety related items
.(i.e.,
Category IIA downgraded to Category 11 components)
-does not-in any way sffect the prenant accident probability while Shoreham is in a defueled status.
The alimination of the tracking requirement to ensure reversibility does not affect in any manner accident consequences while Shoreham is in a defueled status under a Possession Only License.
None of the corrections, clarifidations, or updates.(or the results of these revisions) are involved in the cause of any accident scenario analyzed in DSAR Chapter 15 so they cannot increase the probability ci-occurrence of these accidents.
Relief from the requirement of reversibility (i.e.,
retrievability), which merely involves the tracking of m;difications and part changes, does not have any physical consequences while Shoreham is in a defueled status.
II.
No.
The tracking of modifications and spare parts-for QA Category IIA components was done for future safety (i.e.,
if the plant'ever became operational) an6 not-for-the safety of the plant in the defueled mode.
All systems important to 4
4 safety in the defueled condition remain safety related in-the DSAR.
Revision 3 of the DSAF merely removes the requirement of reversibility from systems no longer considered safety related in the plant's defueled condition.
III. No.
Sen I and II above.
I
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LSNRC-1904 l
l Enclosure A L
Page 17 of 24
-SP,91-011 Description of_ change This generic safety evaluation was performed to provide the basis for determining that facility changes to QA Category II, Seismic Category N.A.
components, systems, and structurec during the effective time of the Posseesion Only License do not present an unreviewed sofnty question.
It-is anticipated that, with the issuance of the Possession Only License (POL), many of the plant activities will result in permanent configuration changes.
The POL-permits activities such as decontamination, component dissambly, and shipment and storage of spent fuel._ Activities to dinmantle and dispose of non-radioactise components and structures not required for safety in the defueled condition or fac1 nandling are permitted.
The POL also permits dismantlement and disposal of radioactive components not required for safety in the shutdown condition, provided that such activity does not involve major structural or other major changes.
This generic safety evaluation is limited to facility changes and DSAR/USAR chunges that are:
Classified as QA Cat. II, Seismic-N.A.
Not associated with the Technical Specifications Not associated with fuel handling or outside temporary ldquid radwaste tanks Not increasing radioactive inventor
- of the site Not resulting _in substantially dif 4erent manner of handling i
or treating fuel and/or radioactive waste and material Not resulting-in increased inventory of nuclear fuel on the site Not affecting the methods or options available for r
decommissioning Not increasing substantially the cost of_ decommissioning Summary I.
No.
The-proposed. changes to the facility and/or DSAR/USAR do pot. alter the plant configuration from that presented in Chapter 15, " Accident Analysis," of the.DSAR.
The proposed changes do not cause-an accident in fuel handlir.g, liquid radwaste tanks or result in piping failures that would create a radiological release.
Chapter 15, " Accident Analysis", of the_DSAR presents the worst case scenario of.-
the consequences of accidents associated with a liquid radwaste tank rupture and a fuel handling accident, the analysis of the consequences of these accidents uhow that offsite doses are significantly less than 10CFRiOO dose limits.
Since there'is no increase in radioactive
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LSNRC-1904 Enclosure A Page 18 of 24 inventory, all of the proponed changen are bounded by the present accident analysis.
The proposed changes will not result in the fuel and radioactive war and material being handled or treated in t nuhutantially different manner than the original bonis of the plant and Possession only License.
The small amount of radioactive waste aml materials are stored in systems and components that were denigned for their control.
The propoced changes do not alter or change equipment important to safety that would increase the probability, consequences or poncibility of a malfunction.
No nafety equipment is required to function during fuel storage in the Spent ruel Storage Pool.
The malfunction of equipmer.t in nupport systems required to be operable during fuel handling operations will recult in auspension of funi handling operations.
The suspencion of fuel handling operations shall not preclude completion of the movement of fuel, equipment or components to a safe conservative ponition.
Therefore, the malfunction of equipment is not critical for the cafety of the plant or its porconnel.
=
II.
No.
See I above.
- ty an defined in the basia for any III. No.
The margin of wa4 technical specification will not be reduced because the structural integrity of the Reactor Building, Spent Fuel Storage Pool, Fuel llandling Equipnent anct components above the Spent ruel Storage Pool shall be maintained.
The failure of any support nyctem or equipment requirei to be operable during fuel handling operations wi.il suspend operations but shall not preclude ccmpletion of the movement of fuel, equipment or components to a safe conservative pouition.
Thorofore, the margin of cafety will not be reduce <1 by any of the proposed changes to the facility or the DSAR/USAR.
a.
LSNRC-1904 Enclosure A Page 19 of 24 4
SE 91-012 Description of Change This safety evaluation was performed to provide the basis for determining that revision 17 to the Offsite Dose Calculation Manual (ODCM) did not involve an unroviewed safety question.
Shoreham's Radiologien1 Effluent Monitoring Program (REMP), as specifit$ in the governing ODCM, was concurrently reduced in scope, reflecting the non-operating and defueled status of the plant. (see also SE 91-009).
Sampling frequencies for certain offluent pathwayn were reduced
-and analysis requirements for isotoper no longer available for i
release were doloted.. Invironmental sampling requirements for milk and ground water were deleted since the pathways do not currently exist.
The outer ring of direct radiation monitoring stations was eliminated since potential exposure from routine Shoreham related radioactivity is insignificant and all dose l
rates are evaluated to be at natural background level at these locations.
Environmental sample analysis requirca.ents were 2
rnduced nimilar_to the. effluent analyuis requirements.
In j
addition, ODCM control and surveillance requiremants for the i
Ventilation Exhaust Treatment System, Containment Purging and Venting and the Gaseous Radwaste Treatment System were eliminated.
The scopo change of revision 17 to the ODCM was identified to the NRC in Section 11, " Miscellaneous Special~ Reports," of the Sem(annual Radioactive Effluent Release Report forwarded by LILCO
_ letter SNRC-1825 dated August 29, 1991.
Summa _r,y.
-1.
No.
ODCM changes are programmatic and do not affect plant configuration or procedures for plant operation aside from providing methods for ke?eping track of and limiting radiation effluents from the plant.
II.
No.
See-1 above.
III. No.
See I above.
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LSURC-1904 Enclosure A i
Page 29 of 24
,SE 91-026 Desc;iption of Change
-Thi s-safety evaluation was prepared to provide the basis for determining that the removal of the Architect Engineer (A/E) of 5
Record for the plant construction, retained to supply the necessary offort to maintain the safety and operability of the plant under existing, modified, or new, approved procedures (see USAR Section 17.2.3), did not involve an unreviewed safety questicn.
In LILCO-letter SNRC-837 dated February 10, 1983, LILCO committed to rutaining Stone and Webster Engineering Corp. (SWEC) for plant
_i design changes of the operating plant until-the first refueling, and to train and-phase in LILCO's own engineering support organization by_the first refueling.
In 1986, LILCO' began a Transfer of Design Responsibility program to systematically transfer identified scopes of work and documents from SWEC to LILCO until responsibility for the complete plant design was transferred to LILCo.
Thcl design responsibility-transfer program was completed in July 1990.
LILCO has developed and implemented procedures and programs with their own qualified personnel to become fully qualified for the engineering and operating of the plant, r
LILCO Engineering and Licensing were the lead organizations in developing the Defueled Safety Analysis Report (DSAR) and Defueled Technical Specifications.
The Posr,ession only License (POL) has changed the design basis-of the plant from an operating plant to a-defueled plant.
The A/E played a very minor role in the program development for the POL.
Based on this, an A/E to maintain the safety and operabilitv cf the plant under existing plant' conditions is no longer required.
In1 addition, all; fuel has been removed from the reactor vessel' and placed in.the Spent Fuel Stor'ge Pool.
Shoreham will never
. approach 1a refueling outage.- Therefore, this commitment is no longer valid based on the current and future plant con. figuration.
' Summary I.
No.
LILCO personnel are qualified to prepare and review the documents required'to maintain the safety and operability of the plant in the defueled condition.
Maintaining a qualified' Architect Engineerlof Record will not add or deter from the safe operation of the defueled' plant.
Since the plant is defueled and never scheduled to operate, the original design basi's has changed and that change waa developed by qualified LILCO personnel.
Therefore the e
1 LSNRC-1904 1:nclosure A Page 21 of 24 reasons for maintaini..g SWEC as an overall reviewer are no longer warranted or valid.
II.
No.
The removal of the A/E of Record does not involve any modification to plant operating procedures or equipment.
The removal of their review does not cause any increase in the possibility or creation of an necident or malfunction.
III. No.
This change.does not impact setpoints or operating margins of the defueled plant.
The basis for Defueled Technical Specifications was developed by qualified LILCO personnel, e
a
. _ _ _ _ _ _ _ _. _ _ ~. _..___
LSNRC-1904 l
Enclosure A Page 22 of 24 SE 91-027
)
i Descrip, tion of Change 1
This safety evaluation was performed to provide the basis for determining that-the segmentation of the drywell shield plugs and modifications to the shield plugs and lifting assembly required to lift the segmented pieces of the chield plugs do not involve an-unreviewed safety quest.on.
Summary I,
- No.
Retaoval_ of the shield plugs from the Reactor Building will not have a significant offect on the seismic response Jof the-Reactor.uilding or any safety related components on the refueling. floor.
The reduction in mass will not affect l
significantly the natural frequencies of the Reactor Building structure.
Additionally, the drywell shield plugo are no longer required for shielding radiation since the reactor' is permanent.ly in_ a non-operhtional status and the radiation levc. is found to be within acceptable limits.
The modificatlon to the existing drywell shield plug lif ting i
_ beam assembly, classified as Q.A. Category II, has been designed in accordance with NUREG 612 and meets all strength requirements necessary to perfo an its function safely.
Additionally, the modified 'ifting assembly will be proof tested to 300% of tho.actua. load it will be lifting.
Although the lifting process will take place in the vicinity of the Spent ruel Storage Pool, but not over it, a potential load drop analysis is not evaluated because the requirementa-of NUREG 612 for single ~ failure proof lifting systems have
~been satisified for the lifting arrangement.
l i
l II.
No.
See I above.
III. No.
See I above.
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l IS!1RC-1904 Enclosure A i
Page 23 of 24 SE 91-029 Description of Change This safety eva3uation was performed to provide the basis for determining that the segmentation of the drywell head and nodifications to the drywell head and lifting assemblies required tu lift the segmented pieces of the drywell head do rmt involve an unt:eviewed safety question.
Summary 1.
t;o.
The drywell head is no longer required as a gas tight membrane for the Primary Containment due to the defueled status of the plant.
Additionally, the drywell head is intended to be removanle and is therefore not required to maintain the seismic or structural integrity of the Primary Containment which has been reclansified as 0.A. Category II in the USAK/DSAR.
Removal of the drywell head from the Reactor Building will not significantly affect the natural frequencies of the Reactor Building structure and therefore will not af fect its coismic recponse or that of any safety related components on the refueling floor.
The )ifing assemblies utilized have been designed and proof tested in accordance with 14UREG 61) and meet all strength requirements necessary to perform the function safely.
Although the lifting process will take place in the vicinity of the Spent Fuel Storage Pool, but not over it, a potential load drop analysis is not evaluated because the requirements of NUREG 612 for single failure proof lifting systems have been satisfied for the lifting arrangements.
II.
14 o See I above.
III. No See I above, l
l.S!11tC-19 0 4 Enclosure A Page 24 of 24 SE 91-031 Dencr_iption of Change This safety evaluation was performed to provide the basis for decermining that the n.odification details for sealing the gaps and K described between the two lower dryer / separator plugs 1;$ Salt, 3,
as " removable" in Figure 3.8.1-3 of tP3 USAR/
and the walls of the dryer /serarator corage pool, tuoreby making the plugs no longer " removable", did not involve an unreviewed safety question.
fjummary I.
No.
Seul plates and concrete grout are being p. ovide d for sealing purposee and as a level baco for the doctmmissioning wet cutting station and do not affact the structural integrity cr seismic. response of the Reactor Building, Spent F'tel Storage Pool, or any other safety related components.
II.
No.
See I above.
III. No. The components being modified are not addressed in the Possession only License Technical Specifications.
In addition, there are no plant equipment or operational requirements being eliminated as a result of those modifications.
I I
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