ML20236B833
| ML20236B833 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 03/14/1989 |
| From: | Leonard J LONG ISLAND LIGHTING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| SNRC-1546, NUDOCS 8903210317 | |
| Download: ML20236B833 (53) | |
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/gg'@ LONG ISLAND LIGHTING COMPANY SHOREHAM NUCLEAR POWER STATION P.O. BOX 618, NORTH COUNTRY RO AD e WADING RIVER N.Y.11792 JOHN D. LEONARD, JR.
VICE PRESIDENT. NUCLEAR OPERATIONS SNRC-1546 MAR 141989 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Submittal of the 10 CFR 50.59 Report for the Period January 1, 1988 through December 31, 1988 Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322 References (1) LILCO letter (J. D. Leonard, Jr.) to NRC (Document Control Desk), SNRC-1515, dated November 18, 1988 (2) LILCO letter (J. D. Leonard, Jr.) to NRC (Document Control Desk), SNRC-1520, dated December 9, 1988 Gentlemen:
In accordance with the requirements of 10 CFR 50.59, this letter transmits Enclosure A, Shoreham Nuclear Power Station 10 CFR 50.59 Report for the period from January 1, 1988 through December 31, 1988. 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. Changes that were made to I the initial test program described in Chapter 14 of the SNPS USAR, under the provision of 10 CFR 50.59 (b) , during the reporting period, have been provided in references 1 and 2, in accordance with License Condition 2.C. (4) of NPF-36.
During this reporting period, the SNPS Emergency Operating Procedures (EOPs) were updated to comply with the recommendations of BWR Owner's Group Emergency Procedure Guidelines, Revision 4.
Enclosure B provides the Station Procedure Change Notices (SPCNs) which were issued to perform this EOP update.
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8903210317 890314 1
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PDR ADOCK 05000322 R PNU
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SNRC-1546 Page 2 The format of this report is as follows:
SM/SPCN No. - In the report, 10 CFR 50.59 items completed during the reporting period are listed by their Station Modification (SM) number or Station Procedure Change Notice (SPCN) number. For convenience of reference, these are each listed separately in ascending order. Associated documents, Design Output Pack &ges (DOPs) and/or Voluntary Change Notices (VCs), are also given vnere applicable.
Description of Change - A brief description of the change, test or experiment autix assed by the change document.
Summary - The safety e valuation determination that the change, test or experiment does not involve an unreviewed safety question pursuant to the three criteria of 10 CFR 50.59 (a) (2) .
Should you require any additional information concerning this submittal, do not hesitate to contact this office.
Very truly yours, e f" x Acleut4L n D. eonard, J .
Vi e President - clear Operations a :ck Attachment cc: S. Brown W. T. Russell F. Crescenzo l
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i j L ENCLOSURE A 10 CFR 50.59 REPORT ~
Period.from January-1,.1988 through' December 31, 1988
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i Shoreham Nuclear Power Station ;
Docket No. 50-322 March 1989 l*.
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Description of Change:
This modification was required to provide a resolution to'TMI ,
Action Plan Item II.E.4.2 for the addition of an' automatic l closure signal to containment purge / vent isolation valves (1T24*A0V001A,B; 1T24*A0V004A,B; 1T46*A0V078A,B; 1T46*A0v079A,B)
,on.a high radiation level in the purge / vent exit flow path. The high radiation alarm signal will originate at the Containment Drywell Filter Train Exhaust Radiation Monitor 1D11-PNL-019 and initiate the Reactor Building Standby Ventilation System. The automatic closure of containment purge / vent isolation valves'will result from this initiation.
All hardware modifications to the plant required by SM 83-140 were completed prior to August 17, 1984-and reported in FSAR, Revision 33, September 1984. SM 83-140 is included in this report since the sensitivity and HIGH' ALARM setpoint for monitor 1D11-PNL-019 reported in the USAR'(FSAR, Rev. 33) have-been modified as follows:
r Old New Sensitivity 5x10 -6 uci/cc, Kr-85 1.05x10 -6 uci/cc, Kr-85 Setpoint 3x10 6cpm 5 5.4x10 crm Summary -
I. No. This. modification provided for the addition of a QA l Category II radiation alarm signal which is not required for q the safe shutdown of the plant.- This signal does perform,a ;
function which will reduce the potential for a radioactive l release via the four (4) and six (6) inch vent / purge valves. l The interface between QA Category I and QA Category II- ,
components is provided by qualified relay coil / contact i separation which will not cause or increase the. potential of 1 a malfunction of any safety related equipment.
II. No. This modification only enhances the ability of the plant to automatically mitigate radioactive releases. A possible malfunction of the high radiation isolation signal will not affect the plant's ability to achieve safe shutdown or to isolate the vent / purge exit flow path through other means.
III. No. The basis for Technical Specifications have not been l diminished, nor have any plant operational requirements been I
eliminated as a result of this modification. This modification provides automatic closure of containment isolation valves which were previously manually closed by the operator upon a high radiation alarm. This enhances the plant's ability to mitigate the consequences of an accident or potential radiation release. :
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Description of Change: .
Radioactive Liquid. Waste System' expansion joints 1G11-EXJ-045, 046, 047,.and 048 were experiencing excessive leakage and, in one case (1G11-EXJ-046), failure. To resolve these problems, the
- expansion joint related pumps (1Gil-P-152 and 157). flanges and.
associated. piping were reworked to allow them-to function as originally designed.
Summary I. No. This modification insures the structural integrity of ,-
the affected expansion. joints and does not adversely affect l their operation or any.other. plant equipment.
II. No. See I above. :
I III. No. See I above.
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~SM 85-015: (DOP.84-128)
Description of Change:
," 'The problem of breached piping due to erosion has occurred
-immediately downstream of1 restriction orifice plates in the radwaste system. . Analysis determined that the following restriction orifices' required modification:
L 620A- 620B 623A 623B 621A 621B 117A ll7B 189A 189B 192A 192B i The modifications made to eliminate the breaching problem were
-threefold. First, where the pressure drop across a restriction !
orifice was calculatedly (or physical evidence indicated it) to be l excessive, two orifices in series were substituted.- Second,-
eccentric orifices were1 replaced with concentrically bored orifices. Third, stainless steel (316L) flanges and pipe'were installed for ten diameters on the downstream side of each restriction orifice.
For other orifices with eccentrically bored plates that were not identified as problematic, only the second and third modifications were implemented.
Summary: 9 I. No. The modification is to non-safety related equipment.
.The modified equipment has no interaction with any safety ,
related equipment. l II. No . - All changes are non-safety related, non-seismic and are not related to any safety related equipment.
III.ENo. No safety related equipment is affected int this changed. The modification has no effect on radiological i L
impacts on the offsite general public, since it does not affect the limits in Technical Specification 3.4.11.1'.
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SM 85-021 (DOP 84-156) l Description of Changes-In Service Information Letter 721-PSM-167.1, April 21, 1982, i General Electric stated that a possible defect exists with Liberty Control Company telephone relays.used in GE protective relays. In accordance with LILCO's response to a NRC Reportable Deficiency 82-00-06, all GE Class 1E protective relays which incorporate the potentially defective relays were replaced.
Eight (8) Class 1E relays were identified as having the suspect date code and six'(6) Class 1Es relays were found with missing date code stickers; all fourteen (14) relays were replaced.
Three (3) of the relays to be replaced were GE type 12NGV25A1F, which are no longer available as qualified Class 1E. These were replaced with qualified 12NGV13B21A relays. ,
l The relays which were replaced are:
Switchgear Bus Relay Type Relay 4160VAC Bus 103 12 SAM 11B22A 62-103-1 12 SAM 11B22A 62-103-2 12 SAM 11B22A 62-1-103 12NGV25AIF 59-103-8 ,
4160VAC Bus 101 12 SAM 11B22A 62-101-1 12 SAM 11B22A 62-101-2 12 SAM 11B22A 62-1-101 12NGV25AIF 59-101-8 4160VAC Bus 102 12 SAM 11B22A 62 ."02-1 f
12 SAM 11B22A 62 102-2 l 12 SAM 11B22A 62-1-102 12NGV25AIF 59-102-8 125VDC Bus Al 12NGV29A2A Gnd. Det. i Bus B1 12NGV29A2A Gnd. Det.
Summary:
I. No. The replacement of the potentially defective protective relays with qualified replacements does not alter the operation of the Class 1E electrical distribution equipment.
1 II. No. This modification does not change the operation of any l of the Class 1E electrical distribution equipment but, rather, increases the reliability.
III. No. This modification in no way prevents the equipment from performing as originally designed and does not change the operating limits of the system.
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- The additionLof the respirator cleaning and drying equipment in ,
the Radwaste Building laundry room required that'the ventilation l system (V41) be modified. 'A respirator cleaning module hood !
exhaust system was connected to the existing laundry dryer >
f (1G11-ELA-002)Ldischarge duct upstream of damper V41-MOD-035. ,
Damper V41-MOD-035 was relocated and existing dryer ductwork- !
modified to permit the operation of either the laundry dryer or the' respirator ~ cleaning module hood exhaust' system. The respirator dryer exhaust system-was connected to the existing laundry room exhaust-system.
Summary:
i I. No. The equipment involved in this modification is not, nor does it interface with, safety related equipment.
II . . No. See I above.
III. No. There is no. applicable technical specification.
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Description of Change: (
Based on the acceptance test, the recirculation loop provided for l
evaporator bottoms tank 1G11-TK-066 was incapable of-maintaining solids'in suspension. To correct this condition, a modification )
- was implemented which consisted of
- (1) the addition of internal J piping to the' evaporator bottoms tank with an arrangement of four l mixing. eductors; (2) the removal of flow control orifice (RO618) I
.with its mounting flanges; and (3) the replacement of the impeller and motor on the regenerant evaporator bottoms pumps '
(lGil-P-154B). The mixing eductor system will impinge on the-tank bottom, thereby causing a swirling motion of the evaporator 4 bottoms liquid. The new impeller and motor are sized to' provide j a sufficient fluid volume and velocity to lift most suspended solids from the tank bottom.
l Summary: )
No. The modification is to non-safety related equipment,
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- 1. 1 powered from a QA Category II power supply. The non-safety j related equipment has no interaction with any safety related i equipment.
II. No. Liquid flow paths have been enhanced thereby decreasing any possibi]ity for different types of accidents or malfunctions.
III. No. The modification enhances the operating characteristics of the evaporator bottoms system. The probability that the evaporator bottoms system could cause an offsite dose in excess of that established in the USAR is unaffected by this modification.
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Description of Change: ?
The originally supplied.and installed suppression pool resistance 1 temperature detectors (RTD) (1Z93*TE112Z, 1Z93*TE113W and 1 1Z93*TE113Z) , manuf actured by Rosemount, Inc., required replacement. A QA Category I RTD manufactured by Conex {
3 Corporation was used as a replacement, as the particular J Rosemount RTD is no longer available. The replacement Conex Corporation RTDs were purchased under SWEC Specification SH1-406A.
Summary:
I. No. The replacement of the original RTD with an equally qualified substitute does not alter or effect in any way the operation of the system in which the RTD is installed.
II. No. No equipment has been added,. deleted or functionally altered by this modification.
III. No. The functional capabilities of the system have not been altered. This modification will have no effect on the radiological exposure to the offsite general public.
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Description of Change:
During surveillance testing on the Intermediate Range Monitoring i (IRM) System it was necessary, in order to adjust the gain of the Amplifier and Attenuator Card, to remove the module from the drawer, open the enclosure, put the module on a test card and i reinsert the assembly back into the drawer. This operation ;
caused the signal-as measured from the amplifier to drift out of '
the calibration tolerance. The problem was due to temperature variations induced in the Amplifier and Attenuator Card when it was removed from the drawer. To. remedy the situation, a 3/16 inch diameter hole was drilled through the side of the module and through the support bracket in-the IRM drawer so that a technician can access the gain adjustment via a screwdriver with the card remaining in place.
Summary:
I. No. This modification has no effect on any equipment.
II. No. This modification has only improved the accessibility for ease of calibration and will not alter the safe operation of the IRM System.
III. No. There is no impact to the Technical Specifications.
This modification will have no adverse affect on the I
radiological impact on the offsite general public.
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Description of Change: '
This modification uses an existing cable between the Station Protection and Metering Panel (1H11-PNL-RU3) and the 13L ,
Switchyard Tie Breaker Control Panel "A", and a new cable between Control Panel "A" and the Control Data Corporation 44.500 Supervisory Panel, to provide the Hicksville Control Room with main transformer breakers (IS22-OCB1310 and 1S22-OCB1330) position indication through the Computer Operated Supervisory System.
To resolve erroneous trips of the sudden pressure relays, which monitor the Main Transformers (S11-T001A and S11-T001B), Westing-house type TD-5 timing relays (62T1A and 62TlB) and Hathaway Trip Current Indication Relays (62T1AX and 62T2BX), located in the Unit Primary Protection Panel (1H11-PNL-RU1) , were disconnected.
Inclusive in the modification was the removal of the 63X and 63V relaying scheme. By removing this portion of the relay protec- l tion scheme, the sudden pressure relay will no longer provide a trip function to the subject breaker control circuits.
Summary:
I. No. The main transformer breaker position indication modification is non-safety related and has no interaction l with any safety related system. Although the removal of the main transformer sudden pressure relay trip permissive from the Unit Primary Protection Scheme is a non-safety related modification, it minimally decreases the probabilities l addressed in USAR Section 15A.1.1, by eliminating a 1 historically erroneously actuated Turbine-Generator trip signal.
II. No. This modification is non-safety related, QA Category II and does not alter the original design bases of the electrical distribution / protective relaying systems. Main transformer faults detected by the sudden pressure relays are alarmed in the Main Control Room, but fault clearing is accomplished by existing protective relaying.
III. No. No safety related equipment is affected by this change, and this modification in no way prevents the Main Transformers from performing as originally designed.
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Description of Change:
The object'of this station modification was the installation of a motor-driven trash rake assembly to facilitate periodic cleaning l of the screenwell trash rake bars at each intake bay. 1 Previously, marine _ growth and miscellaneous debris was manually I removed which proved to be ineffective in controlling accumulated '
debris. The design change'is considered non-safety related and QA Category II. Anchor bolt holes were drilled in the screenwell conc _ete structure which is considered QA Category I, seismic Category I and safety related.
Summary:
I. No. The modification has no direct effect on the screenwell i structural integrity or safety related components located in j the screenwell. The trash rake, when in the submerged position, is external to the trash bars and does not impact the operation of any safety related pumps or components.
II. No. The installation of the trash rake is non-safety related and QA Category II. The ultimate heat sink will ,
always be available since the trash rake operates in one bay !
at a time and malfunctioning in the submerged position will j not render all intake bays inoperable. !
III. No. The installed equipment does not affect any Technical l Specification related equipment which could concern plant operability.
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Description of Change:
In order toLimplement'the tie-in of the Colt emergency diesel 1 generators (EDGs) with the existing TDI EDGs, a new 4 KV-switchgear_ cubicle:will;be required in Emergency.Switchgear Room
.103. This modification provided a new sill for the switchgear cubicle. Also, new sleeves were added to' provide a means of ;
routing power,' control, and-instrument cables required for the. 1 implementation of the Colt tie-in.- The sleeves will be spare ;
-until:the installation of raceways.and cables.
Summary I. No. The modification does not effect the seismic integrity of the~ Control: Building, qualification of the blockwalls, or the structural integrity of the Control Building structure.
The fire rating of the. walls is not affected.
II. No. The additional' sleeves and blockouts do not decrease the seismic. integrity of the. block wall or the structural integrity of the Control Building. The sleeves were sealed so as not to: decrease the fire rating of the walls, i i
III. No. This modification does not affect any safety related equipment.
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Description of Change:
Each of the three emergency diesel generator's (EDGs) independent fuel oil. supply systems will be modified to supply fuel from one l fuel oil storage tank and the associated supplementary tank to l
either one Colt.EDG or one TDI EDG, as desired. The scope of SM-85-066 is' limited to initial modification within the fuel oil transfer pumphouse for installing the necessary valves and piping components which directly interface with the existing fuel oil l supply system. The discharge piping for pumps 1R43*P-202A, B, l and C were modified to. accommodate a manual valve and a ,
tee-branch connection for future tie-in to the fuel oil. piping ;
for the Colt EDG. Presently, the branch connection connects to a ;
short length of pipe which terminates at a blind flange. For operation with the TDI EDGs only-(i.e., prior to the Colt EDG tie-in) , . the new manual valve will be open. Upon completion of the~ Colt EDG tie-in, this valve will be closed for isolation of the cross-tie between the Colt and the TDI EDG fuel supply subsystems.
Also, under SM 85-066, a manual valve was added to the existing common discharge piping inside the fuel oil pumphouse.
Presently, this valve will be open for supplying fuel to the TDI EDG. Upon completion of the Colt EDG tie-in, this valve will remain open, but can be closed for diverting fuel oil to the day tank for the Colt EDG when required.
Summary I. No. The operation of the fuel oil system remains unchanged by this modification, and the design, fabrication, and installation of this modification are consistent with the i remainder of the system.
II. No. This modification does not alter the design basis of the Fuel Oil' System. Pipe and valve segments were fabricated, installed, and tested consistent with the remainder of the Fuel Oil System. The piping is seismically supported and the concrete vault provides missile protection.
III. No. This modification does not reduce the' capability of the Fuel Oil System to transfer fuel oil to the respective diesel-generator day tank. In the event that the valve on the discharge of the fuel oil transfer pump is shut, the upstream components will remain protected by the fuel oil i transfer pump discharge relief valve. Piping components were fabricated, installed, and tested consistent with the remainder of the Fuel Oil System.
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' Description of' Change:.
As part of the Colt emergency diesel generator tie-in, new raceways and supports are: required for the routing of cables in the Relay Room, HVAC/ Chiller Rooms, and-TDI. Rooms. The new '
raceways and supports were added'according to the. design =
criteria, including seismic, to provide a means for routing
_ power,. control and instrumentation cables required for the implementation of the Colt tie-in.
Summary I. :No . - The addition of raceway and supports, and the attachment to' existing supports, have been analyzed for seismic conditions,'and they meet the design requirements.
II. No. See I above. .
III. No. See I above.
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. . 1 s 1 SM 85-068 (DOP 85-086) l Description of Change:
As.part of.the Colt emergency diesel' generator tie-in, new raceways and supports are required for the routing of cables in Corridor #9, SWGR Room 101, SWGR Room 102, and SWGR Room 103. 4
'The new raceways and supports were added according to the design criteria, _ including seismic, to provide a means for routing power, control and instrumentation cables required for the 1
. implementation of the Colt tie-in. I Summary I. No. The addition of raceway and supports, and the attachment to existing supporta, have been analyzed for i seismic conditions, and they meet the design requirements.
II. No. See I above.
III. No. See I above. ,
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Description of Change:
Inductive " kick" in the electro hydraulic control (EHC) system
. pressure setpoint increase / decrease circuit caused the main turbine bypass valves to operate erratically when'the control Room operator mado EHC pressure setpoint changes. To alleviate this problem and to allow the turbine bypass valves to function l i
as' designed, suppression diodes were added in parallel with relay. !
coils XK15 and XK35 of the pressure control logic circuitry.
Summary:
I. No. No design function of any safety related or non-safety related equipment has been altered. The diodes and the -
wiring added were to QA Category II devices and circuits. '
II. No.. The failure modes of the EHC system have not been altered by the addition of diodes in the pressure control circuitry.
III. No. No design functions, setpoints or automatic actions are altered by this modification, nor does this modification l inhibit any automatic or manual actions. !
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Description of Change:.
As identified in Shoreham's'" Environmental Qualification Report
.for Class lE. Equipment," Rev. 5,. August 1982 (EQR) , eighteen (18) ,
electrically operated valve. actuators located.in a " harsh
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environment" were not1 environmentally qualified in accordance .
with1NUREG-0588-and=IEEE' Standard 323-1974. EQR,' Appendix H, )
" Justification for Interim Operation," provided a detailed l' Janalysis of this. equipment's operability requirements and failure consequences. A test' program was completed.and confirmedithat replacement actuators.with dry. type motor start capacitors are -1
. environmentally qualified. LTherefore, the existing paper: oil motor start capacitors were replaced with the newly qualified. dry type capacitors.
In addition, all " mild environment" and " spare" actuator motor capacitors were replaced with the dry type replacement-motor start capacitors. nThe " mild environment" actuator motor paper oil capacitors were replaced since the oil in these capacitors was PCB based.
I All modifications to the plant required by.this DOP were completed prior to. November'23, 1985.
. Summary:
s I.. ENo. Replacement o'f unqualified capacitors with qualified. ~I capacitors minimizes the potential risk of possible actuator misoperation when exposed to radiation.
II. No. This modification did not change the mode of operation or any' operating requirements, such as external control circuits.
III. No. The modification required that the unqualified motor start capacitors be replaced with equivalent but environmentally qualified capacitors, thus enhancing the capability of the actuators to operate as required under .l normal or accident conditions.
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-Description of' Change:
Circulating. water pumps.1N71-P063 A thru D receive-bearing
- cooling. water from the. Service Water System. Removal of .
- suspended solids from service water is accomplished via parallel M' '
' strainers 1N71-S081A and B. These strainers are automatic, self-cleaning. units with backwash capabilities provided by Zurn Industries (Model No. 595). The materials of construction includedia ductile iron body and cover and a Monel basket. '
L Component. failure resulting from severe' erosion / corrosion
-required that material better suited to a service water
- environment-be: utilized.
Under the modification, the strainers have been replaced with Zurn; Industries strainers'that use a cast type 316 stainless steel body and cover, with the balance of all wetted parts also being 316 stainless steel.
Summary:
I. -No. No safety-related equipment is involved in this change
.and there is no interaction with any safety-related system.
II. 'No. See-I above.
III. No. Strainers 1N71-S081A and B are not discussed in the Technical Specifications.
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[ SM-87-005' (DOP 83-029) . Description of Change ) ) This modification added common fire suppression carbon dioxide predischarge alarm horns.to the.following nine (9) areas: Relay Room, Computer Room, Emergency Switchgear Rooms 101, 102 and 103, HVAC Equipinent' Room - el. 44 ft.-0 in, Chiller Room - el. 44 ft - 0 in, corridor between Emergency Switchgear_ Rooms and-Battery' ) Rooms A and B, and North Stairwell in the Contr61 Building. The ? common alarm horns will ensure the safe evacuation of personnel , from these areas prior to the discharge of CO ' 2 Summary: ' I. No. No safety related equipment is affected by this change. II. No. No safety related equipment is affected.in( this change. III. No. The technical. specification is unchanged. i i A - 20 ~ - - _ - - _ _ _ _ _ _ - _ _ - - _ _- - ___ - _ i .3 L ... . . SM 87-013 (DOP 86-109) Description of Change: In order to decrease the potential for particulate plateout in sample lines for radiation monitor panel 1D11*PNL-021 (low range post accident station ventilation exhaust) and 1D11*PNL-022 (Reactor Building Standby Ventilation System) and to reduce personnel exposure during changeout of the post accident sample filters (1D11*FL100 and 1D11*FL101 in the sample lines to the above mentioned panels), the following changes were made: 1
- 1) Isokinetic nozzle 1D11-RN141 (providing sample to 1D11*PNL-021) was modified so that it exits the station vent duct on the south side instead of the north. The existing
, sample line was routed through a branch line designated 2" l SM-300-153-3 to 2" SM-23-153-3, which then goes to 1D11*FL101. This branch line was deleted and 2" SM-23-153-3 run directly from the nozzle. The existing blanked off and retired in place section of 2" SM-23-153-3 going to nozzle 1D11*RN-021 (in the RBSVS) was removed, including 1D11*RN-021.
- 2) Sample filters 1D11*FL100 and 1D11*FL101 (located in. sample lines to PNL-022 and PNL-021, respectively) were replaced with filters incorporating a quick disconnect feature.
- 3) Machined concentric reducers and reducer bushings upstream of the above filters were replaced with reducers having a more gradual change in inside diameter.
- 4) Heat tracing was re-worked to suit the.new routing-of 2" SM-23-153-3.
Summary:. I. No. This line is not analyzed for any accident in the USAR, and this modification involved only minor piping changes to provide a more representative sample of particulate. II. No. This modification improved the performance of the monitor, and did not add any new safety related equipment. III. No. This modification does not change what is released to the atmosphere, but allows better monitoring of releases while reducing personnel exposures. l l I A - 21 1 SM 87-019 (DOP 84-133) Description of Change Effective carbon dioxide fire suppression'in the normal switchgear room requires that the room's rollup steel door be fully closed. SM 87-019 changed the door's control circuitry ' such that a person need only momentarily depress the "close" push button to fully close the door. Previously, the control circuitry required that a person keep the door's "close" push button depressed until the door was fully closed. In the event of a fire, a person would have had to remain within the area and be unnecessarily exposed to a CO hazard. 2 Summary: I. No. The change made to the control circuit has no effect on any safety related systems, components or structures. II. No. The failure modes of the new control circuitry are , identical to the failure modes of the original system. l III. No. No setpoints or design functions of any systems, components or. structures were altered by this modification. This modification has no impact on radiological exposure to the offsite general public. A - 22 SM 87-032- (DOP 85-243) Description of Change Balanced magnetic switches (BMS) were installed on various doors throughout the plant to ensure that the fire protection systems / areas they surround are closed. The BMS's were wired to the continuously monitored security computer system. A trouble alarm will be initiated at the security consoles when these doors are open for longer than a prospecified time. This modification eliminated the fire watches previously used for these' areas. Summary: I. No. No safety related equipment is involved in this modification. II. No. This modification involves only non-safety related components. III. No. See I and II above. l l l l A - 23 L_____________-____________ \- SM 87-033 (DOP 86-007) i Description of Change: i The external alarm circuit of the Loose-Parts Monitoring System was generating random alert and reset signals in the monitoring i circuit. This was due to the inductive " kick" caused when relays in the' external alarm circuit were de-energized by alarm acknowledgement. The condition was corrected by isolating the external alarm circuit from the monitoring circuit by utilizing a small power relay to energize /de-energize the external alarm circuit. Summary: J I. No. The Loose-Parts' Monitoring System is non-safety related QA Category II. The modification does not affect, and has no interaction.with, any safety related equipment. II. No. All changes were non-safety'related. III. No. No safetyLrelated equipment is affected by this change and there is no impact on the Technical Specifications. The modification has no effect on radiological impacts on the offsite general public. h I t ) A - 24 3 . '. .t SM 87-035 (DOP 85-177) ' ' Description of Change: RCIC INOP alarm window (GE Alarm 1083) did not light when the outboard warmup bypass valve (1E51*MOV048) was not fully open. This was contrary to a dedicated alarm (GE Alarm 1086) displaying the valve not-fully.open. Limit switch contact #11 on valve (1E51*MOV048),' which provides the dedicated alarm for "not fully open", was multiplied by an auxiliary relay. One contact of this relay was_used for the dedicated alarm (GE Alarm 1086), the l second contact was inserted into the inoperative alarm string, ' and the third contact was used for the computer information system. This modification was performed in compliance with GE ; FDDR KS1-2361, Rev. O, December 11, 1985 and Rev. 1, October 6, I 1986. i Summary: I. No. This modification provides additional'information helpful to the operator and does not change any safety ' functions. A QA Category II auxiliary relay was installed in a QA Category II panel in the Relay Room. The existing and new cables used for this modification are QA Category I ' and the raceways are qualified Seismic I. ) II. No. See I above. ! III. No. See I above. j ( i l A - 25 SM 87-036 (DOP 85-168) Description of Change: HPCI INOP alarm window (GE Alarm 1045) did not light when the outboard warmup bypass valve (1E41*MOV048) was not fully open. This was contrary to a dedicated alarm (GE Alarm 1047) displaying the valve not fully open. Limit switch contact #11 on valve (1E51*MOV048), which provides the dedicated alarm for "not fully open", was multiplied by an auxiliary relay. One contact of tb's relay was used for the dedicated alarm (GE Alarm 1047), the second contact was inserted into the inoperative alarm string, and the third contact was used for the computer information system. This modification was performed in compliance with GE FDDR KS1-2362, Rev. O, December 11, 1985 and Rev. 1, October 6, 1986. Summary: I. No. This modification provides additional information helpful to the operator and does not change any safety functions. A OA Category II auxiliary relay was installed in a QA Category II panel in the Relay Room. The existing and new cables used for this modification are QA Category I and the raceways are qualified Seismic I. II. No. See I above. III. No. See I above. l A - 26 L_____----__----- - 4 p- , , .,
Description of Changer, Piping supports for'the non-safety related, QA Category II, feed-water piping system . (IN21) were modified to accommodate addition- j al loads due to thermal stratification. This modification was initiated in response to NRC Information Notice 84-87, " Piping Thermal Deflection Induced by Stratified Flow." The following pipe supports were modified: ) IN21-PSST 418 1N21-PSA459 1N21-PSST 421 1N21-PSSH460 1N21-PSST 438 1N21-PSR461 : IN21-PSR439 1N21-PSR465 IN21-PSR441 1N21-PSR510 IN21-PSR442 1N21-PSR552 1N21-PSA443 1N21-PSSH1800 1N21-PSSH444 1N21-PSSH1801 l 1N21-PSR445 i i Summary: I. No. This modification corrected potential effects associated with thermal stratification. t II. No. The existing functions and purposes of the piping system remain unchanged. III. No. This modification does not effect the functions of the piping system to achieve cold shutdown. 1 i i L . A - 27 - __--__--_-_-___-__-__-__ . _ . t I . .. j .- SM 87-041' (DOP 86-111) < Description of Change: 1 Radiological Environmental Monitoring Program (REMP) air samplers l (3S1 and 6S2) were experiencing low voltage at their motors caused by undersized cables. The air samplers are located on Beach Road east of the plant, over 500 feet from the power panel ; (1R35-PNL-N15). New larger power cables and underground conduit > -were routed from the power panel to the air samplers. The old power cables were abandoned and kept as spares. l Summary: I. No. The modification is to non-safety related QA Category II equipment. The modification did not affect and had no interaction with any safety related equipment. II. No.~ All changes are non-safety related and are not related to any safety related equipment. III. No. No safety related equipment is affected by this change and there is no impact on the Technical Specifications. The modification has no effect on radiological impacts on the offsite general public. A - 28 1 . > i .o SM 87-044 (DOP 87-074) j Description of Change:. A review, prompted by I.E. Information Notice 83-08, determined that Square D Type DO-22 relays were not rated for Shoreham's control batteries voltage levels. This modification replaced all seven (7) existing Square D Type DO-22 relays with Agastat EGFD l 002 relays which have a rated coil range of 100 VDC to 140 VDC, suitable for the Shoreham DC voltage levels. The three relays (74-1R23A05, CO6, and C07) in the shunt trip panels 1R23*PNL-001 and 003 are QA Category I. The four relays (23Y-1U41A03, B03, A04 and B04) in panel lH21-PNL-VX5 are OA Category II. Summary: I. No. This modification only replaced relays that had inadequate coil voltage ratings with relays that have adequate coil ratings. Therefore, the probability of relay failure is reduced as a result of this modification. II. No. The new relays have a higher coil voltage rating than the previous relays with a corresponding lower possibility of failure. ; III. No. The . functional capabilities of the system have not been altered. This modification has no effect on the radiological impact to the offsite general public. ll A - 29 l L i L 1 f ! L -. . , p
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SM 87-046 (DOP 85-221) l i Description of Change: I L The object of-this modification was to tie-in, test and make f . operational the Security System in the Colt Building. Also, as a result of the construction of the new Colt. Electrical Cable Vault and Access Structure, security components from door C-25-1 were- , relocated'to new door.EV-25-1. The Security System is non-safety l related. l Summary I. No. This modification did not affect any equipment required I for the safe shutdown of the plant.. II. No. This modification has a design basis equivalent to that l described in the Security Plan referenced in the USAR. The i equipment added and/or relocated by this modification-provides no safety related function. III. No. No plant operation requirements have been revised or I added. The Station Security System is not addressed in the l Technical Specifications. l q i l l A - 30 l .. l l SM 87-047 (DOP N/A) Description of Change: Past failures of suppression pool resistance temperature detec-tors (RTDs) led to a design change which eliminated the cable to lead termination at the RTD heads. This was accomplished by replacing the existing RTDs manufactured by Rosemount with RTDs , supplied with 100 feet of lead wire manufactured by Conax. Be- I cause this new configuration eliminates one of the terminations inside the suppression pool, it eliminates the most common , problem area. The new RTDs required that the conduit connections l at the RTDs.by modified slightly. The scope of this modification ' was limited to the following twelve (12) RTDs, as described in E&DCR L-1228C: l 1Z93*TE110W, X, Y, and Z 1Z93*TE112W, X, Y and Z* 1Z93*TE132A and B 1Z93*TE134A and B
- The original Rosemount manufactured RTD (1Z93*TE112Z) was replaced by a Conex manufactured RTD under DOP 85-026 and implemented from E&DCR L-0884C. The Conex RTD was replaced by this modification.
Summary: I. No. This modification is an upgrade in the element to wiring termination design and therefore increases the component's reliability which effectively reduces the possibility of a safety related component failure. II. No, This modification does not change the design or configuration of those systems in the Accident Analysis Chapter of the USAR. III. No. This modification does not change the design or configuration of those systems described in the Technical Specifications. A - 31 = . SM 87-049 'Descript' ion of Change: Technical Specification'4.11.2.8.3 requires that after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation, a carbon sample be taken from 1T41-FLT-003 and analyzed. In order to accurately tell when i the adsorber has operated for 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />, this modification installed an elapsed time meter for primary containment purge i filter fan 1T41-FN-64,in 1R24-MCC-11D2. 1 Summary: I. No . . No safety related equipment was involved in this design change. l l II. No. This modification involved only non-safety related instrumentation. III. No. No safety related equipment was functionally affected i by the addition of this timer. I ( i l A - 32 s , j ,s-ISM 88-001 (DOP N/A) 1 Description of Change: l The objective of this modification to the hypochlorination sample line was to reduce the potential for bio-fouling, simplify 7 s cleaning the sample. pipe-and eliminate air in-leakage. Since copp$r.is toxic to marine life, the 4" OD x 4" deep sample well was' lined with 1/8" thick copper. A " Pipe Within a Pipe" design was implemented to facilitate removal of IP33-1-1/4"-SM-040-167R j f for cleaning during circulating water system outages. Finally, to prevent vibrational damage, new support IP33-PSR9608 was j l fabricated and installed (see ECR H-00603A and H-00603B). Summary: I. No. The replacement of the sample line does not affect or , change in any way, the operation of the system and does not ' interact with any safety related equipment. II. No. This modification does not challenge the integrity of this system or any other system. The " Pipe Within a Pipe" sample line is functionally interchangeable with the previous sample line and the addition of the new pipe support provides additional protection against damage due to vibration. III. 1k3. .The hypochlorination sample line system is not addressed in any Technical Specification. j . o l l l A - 33 ! SM 88-003 (DOP 86-152) Description of Change: Regardless of what calibration frequency was used, RCIC pressure switches lE51*PS023A and C would not retain calibration. These switches were replaced with switches of the same type but with a different calibration range. The new switches have seismic ! responses such that they could not be mounted on the same panel and, therefore, a separate stand was built adjacent to the panel to dampen the response to within acceptable limits. Summary: I. No. This change provides reliable inputs to the system's logic. II. No. This minor change was designed with consideration of existing design specifications which include separation criteria, scismic integrity, component failures, etc. III. No. This change was not a change to any design basis which forms the margin of safety. This was a minor change which functionally was a "like for like", with other attributes that were improvements. A - 34 1. 1 SPCN 87-1294 Description of Change: i This SPCN added the abnormal performance section, and related appendices, to SP 87.307.10,'" Colt Diesel Generators," for ) starting a unit to a dead bus to backfeed the main plant RSST via l -the 69 KV yard. This capability is a 25% power operation license , commitment. Summary i i I. No. The modification provides an additional means of supplying power during an accident or malfunction. II. No . . See.I above. III. No. See I above. i l 1 I A - 35 m' !- ..., ,- \ , l
- s. . SPCif ' 87-1963 ,
>/' . ) Description of Change: I This SPCN created SP 87.136.01, "SRV Short Term Accumulator Pressure Drop' Test",'as part of the Inservice Test Program. This procedure was committed to in SNRC-0638 and again in SNRC-1153. Summary I. No. Since this' test will only be performed while in ) , operating-conditions 4 or 5 and the Automatic ; Depressurization System (ADS) is only required when the 1 reactor pressure is greater than 113 psig, this test will not impact ADS operability. II. No. See I above. III.'No. See I above. f i 1 j i i A - 36 i _ - _ _ _ _ . . _ - _ _ _ - _ _ _ - _ . _ _ - _ _ _ _ _ _ _ - _ ._ _. __ __- - __ __ _-_A t-p i SPCN 88-0004 Description of Change:. ! This SPCN created SP F4.505.02, " Inspection of Fire Rated Assemblies and Functional Testing of Fire Dampers," from existing SP 34.505.02.- Section 8.4'provides for the inspection of.the q cable tray fire barrier on El 8' of the Reactor Building. This ! is a partial height (9') barrier constructed of 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated material (front) and 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated material (sides) installed between the vertical cable trays and the adjacent RCIC pump. This fire barrier is discussed in detail in the response to NRC u Question 12 of the FHAR and in SNRC-1337. l Summary I. No. Fire loading has been. calculated and a fire on either side of the barrier has been evaluated. The classification of this barrier as a mixed I hour /3 hour barrier remains conservative. Shutdown of the reactor with either IIPCI or RCIC (under the single failure criterion) has been evaluated and does not change because of this modification. II. No. The FHAR calculations for fire hazards has been developed assuming the current configurations. III. No. The barrier will be inspected and considered a fire barrier, and all actions taken in accordance with Technical ! Specifications. 1 .i 4 A - 37 _______----,--__---_-----_----.,-------_.s - - - _ - - - - - - - - - - - - - i SPCN 88-0038 (TPC 88-005) Description of Change: The object of SPCN 88-0038 was to incorporate into SP 87.307.08, " Colt Diesel Generator and Support Systems Line-up", the changes brought about by the extension of the fire protection system boundary to include the Colt Diesel Building. The physical extension of the system was performed under DCN 84-083-8 (E&DCR L-0363N) and will have the capability of pressurizing the fire protection water header up to the deluge valves, hose stations, pressurizing the supervisory air systems, and charging the Halon 1301 system. No active credit is being taken for this modification with respect to 10 CFR 50, Appendix R until the Colt Diesel Building and diesel generators are incorporated into the plant system. SPCN 88-0038 provided instructions for the line-up and operation of Colt fire protection systems, with associated Alarm Response Procedures (ARPs). Summary: I. No. The extension of the fire protection system boundary does not impact system operation but increases overall station fire protection. II. No. The piping is designed and installed consistent with the remainder of the fire protection system. III. No. See I and II above. A - 38 SPCN 88-0211 Description of Change: The subject SPCN provides steps to SP 23.714.01, " Gaseous Radwaste (Holdup)", to prevent the-automatic swap-6Ver feature on the Offgas System. This feature was originally intended to compensate for the fact that the offgas panel is not in.the main control room and that the radwaste control room would not be permanently staffed. Since the SNPS radwaste control room is permanently and continually staffed, automatic swap-over is not-necessary. The present Alarm Response. Procedures (ARPs) properly guide the operator to switchover to the standby train, as necessary. Summary: I. No . - The same functions will be performed manually with the revised procedure, as were previously performed automatically.- II. No. See I above. III. No. See I above. i f 1 l 1
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'SPCN 88-0237 (related to TPC 87-227) . Description of' Change: SPCN 88-0237 SP 12.012.01' 2 !- " Radiation Work Pergit"'from 500 dpm/100 cmchanged contamination . limi 'The 1000 dpm/100;cm limit is recommended by the NRC (IEC 81-07) and INPO.- TPC 87-227 added a step to SP 12.012.01 to give RWP approval authority to the RPC/RAC in the TSC under declared emergency conditions, as well as the Watch Engineer. Summary I. No. The' changes bring the procedure into agreement with NRC and INPO recommendations. Providing the RPC/RAC with RWP approved authority ensures that RWPs will be issued in a controlled but timely manner under emergency conditions. II. No. These changes do not create the possibility for any-accidents / malfunctions not described in the USAR. III. No. Neither of the changed items are discussed in the Technical Specifications. j l A - 40 L. 1 ! SPCN 88-0263 (related to SPCN 88-1615) l l '\ l Description of Change: 1 SPCN 88-0263 modified SP 47.106.03, "ATWS/ARI and Backup Scram , Valve Functional Test", to perform the required functional testing of the ARI (Alternate Rod Insertion) and Backup Scram solenoid valves: 1C11*SOV048A and B (F160A and B), and 1C11*SOV42A and B (F110A and B) and Backup Scram initiating I circuit. The ATWS/ARI function was added, as a redundant back-up system to the Reactor Protection System (RPS) by SM 87-007. A 10 i CFR 50.59 report of SM 87-007 was provided in SNRC-1397, December 4, 1987. Summary: I. No. Functional testing of ATWS/ARI will help to ensure the reliability of the RPS. II. No. Actual rod motion is inhibited during testing by this procedure. III. No. SPCN 88-0263 does not affect the limits of operation of any safety related system. I i A - 41 1 . . i I SPCN Ed-0761 j Description of Change: This SPCN created new SP 87.114.01, " Isolate Phase Bus Air Leakage and Pressure Test", to verify that the air-tightness of the main generator isolated phase' bus duct is within manufacturer recommended limits.- The test will be conducted with the main turbine-generator tripped and the isolated phase bus (IPB) i electrically isolated from the grid. The IPB is not safety related, it is described in USAR Sections 1.2.2.6.1 and 8.2.1. Summary: I. No. The test will be conducted only with the main turbine-generator tripped and the IPB electrically isolated from the grid. II. No. See I above. III. No. The IPB is not mentioned in the Technical Specifications. 1 ( i i t l l \ l' l A - 42 c... 1 p SPCN 88-0805, ! i Description of Change: This SPCN created new SP 47.127.23, "Steau Valve Stroking", to l test the operation of the portion of the Turbine Control System (N32) which controls stroking of the main stop valves, control valves, combined intermediate valves, bypass valves, and the . reactor feed pump turbine stop and control valves. The objective I of this procedure is to functionally test the various valve positioning circuits and effect minor adjustments, if necessary. j i Summary: ' I. No. The plant will not be operating during the performance of this procedurc (mode switch in condition 4 or 5). Also, the equipment will be functioning within its design operating limits. II. No. See I above. ; III. No. See I above. r l A - 43 m a .' - ' SPCN 88-0927 ; I . Description of Change: I SPCN 88-0927 modified SP 87.307.08, " Colt Diesel Generator and Support Systems Line-up", to allow two (2) diesels to be run j simultaneously for' summer peaking. The Colt diesel tie-in a described in SP 87.307.08 is temporary and will be removed. 'j Summary: I. No. The Colt diesel generators are independent of the 'l plant, and associated plant support systems are non-safety related, not related to the Technical Specifications, and are not required for safe shutdown. In addition, protective devices, control circuits, and alarms sufficiently protect the Colt generators and associated equipment, and operation in the two Colt diesel peaking configuration will not' l degrade the category I qualification of the equipment. ! II. No. The Colt diesel generators are independent of the plant, and the associated plant support systems are ! non-safety related. The Colt diesel generator protective devices and control circuits adequately isolate the diesels from the emergency buses such that any fault associated with two diesel peaking will not adversely affect the offsite AC power sources for the plant. Two diesel peaking operation cannot be performed with the plant in' operating conditions i 1, 2 or 3. l III. No. The Colt diesel generators are independent of the ; plant, and the associated plant support systems are non-safety related. The two diesel peaking configuration is limited to plant operating' conditions 4 or 5, and is adequately isolated from the plant emergency buses such that the offsite AC power sources for the plant will not be adversely affected. A - 44 e_-__-___________---____ f- " . p: m , 'SPCN 88-0928' Description of Changes
- l. SPCN 88-0928 modified SP 87.'307.10, " Colt Diesel' Generators", to allow two (2) diesels to be run simultaneously for summer peaking.
Summary: -I. No.-. The_ Colt diesel generators-are. independent of the . plant, and associated plant support' systems are non-safety related, not related~to the Technical Specifications, and . ! .are not required-for safe shutdown. In addition, protective devices, control circuits, and alarms sufficiently protect the' Colt generators and associated ~ equipment, and operation in the two Colt diesel peaking configuration will not degrade the Category I qualification of the equipment. I II. No._The Colt diesel generators are independent of the. plant, and the associated' plant support systems are non-safety related. The Colt-diesel generator protective devices and control circuits adequately isolate the diesels from=the emergency buses such that any. fault ansociated with two diesel peaking will not adversely. affect the_offsite AC power sources for the plant. Two diesel peaking operation cannot be performed with the plant in operating conditions 1, 2 or 3. III. No. The Colt diesel generators.are independent of the plant, and the associated plant support systems are non-safety related. The two diesel peaking configuration is limited to plant operating conditions 4 or 5, and is-adequately isolated from the plant emergency buses such that the offsite AC power sources for the plant will not be l adversely affected. 1 i A - 45 L . SPCN 88-1229 Description of Change: 3 I I This SPCN modified SP 12.075.01, " Administration of Startup Testing," to incorporate.the recommendations of General l -Electric's FDDR KSI-2459. FDDR KSI-2459 expanded Test Condition ] 4 (TC-4) test . window from 100-95% rod lines to.100-75% rod lines, f The expansion of the TC-4 window allows reactor operation in the i region of increased core thermal hydraulic stability. A safety l evaluation of this change was provided to the NRC in SNRC-1520, l December 9, 1988. j l Summary: I I. No. This modification helps to ensure that the SNPS procedures reflect the latest recommendations of General Electric. II. No. See I above. III. No. This modification is consistent with Technical Specification 3.4.1.lb requirements. l l A - 46 l l .e .j TPC 88-71 ! l Description of Change: TPC 88-71 changed SP. 23.420.01, " Main Chilled Water System" by modifying valve 1M60-10V-0278 to be " LOCKED CLOSE" on the valve lineup checklist. This change was necessary due to the system . modification-(E&DCR H-00642A) which upgraded the operating , pressure on the condensing water side and removed and isolated. the rupture disk. i Summary: I. No. The Main Chilled Water System is not safety related. Interface with the safety related Reactor Building Service Water System is at isolation valves P41*MOV-36A, B and C and locked closed valve P41*10V-0031. The MOVs automatically close on an accident-signal to isolate the RBSW from the ! Main Chilled Water System chillers. II. No. See I above. III. No. See I above. l l t l l A - 47 b, t e_., .' . 2 --j.,, ., 4. 'g, ~~l l', ' ' f '. c l['f ) 'h g' yf.'k_i_'; 1 1 1 , . .&
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, ,l,:lg '. il * ' ' l .j < .g:: I 1 r .1, 3 't l { t 5 ENCLOSURE B Implementation!of BWR Owner's Group - Emergency l Procedure Guidelines, Revision 4 l .. l .; ; .\ 1 Shoreham Nuclear Power Station Docket No. 50-322 March 1989 ~ F H ' 'l ~ } J$ ~ e . 8 SPCN LSP ' Title -( 87-0750 '29.020.01 AlternateJShutdown Cooling- j 88-0297l 25.001'.01~ Emergency Operating Procedure. Supplement' 88-0519. 29.023.02 Secondary Containment. Control ; 88-0526" 129.023.01 lRPV Control' " '88-0531 > 29.023.03_ Primary Containment Control .88-0532 .29.023.06 Radioactivity Release. Control 88-0534 29.023.09 RPV Flooding _ ' 88-0535 29.023.08 Primary Containment Flooding ' 88-0536 29.023.04 Alternate Level Control 88-0537' 29.023.07 Level / Power Control Description of Change: j <l ~The above SPCN's either revised or created.the associated 'l emergency operating procedures (EOPs). to comply with the recommendations of the BWR Owner's Group Emergency Procedure Guidelines . (EPGs) , . Revision 4 (EPG-Rev. 11). The EPGs are designed to provide guidance for operator response for all- :! emergency. situations not just the design basis accidents / malfunctions of the USAR. Small-break LOCA, 'large-break.LOCA, transients with multiple failures or no j . failures, inadequate core cooling, and anticipated transients without scram (ATWS)'are all addressed by the EPGs. 'i 1The NRC has performed a safety evaluation of EPG-Rev. 4.. The ~I safety evaluation found the actions specified in EPG-Rev. 4 to be ti " generally. correct and appropriate and within the operators' ; capability." They also determined.that the guidelines were- l acceptable for implementation, and recommended that EOPs be H revised to reflect EPG-Rev. 4 "as early ar practical." The SNPS EOPs were; updated to reflect EPG-Rev. 4 in anticipation of the NRC's approval of the BWROG's submittal. Revision 2 to .} l the USAR revised the' listing of'EOPs provided on Table 13.5.1-1 to reflect the EPG-Rev. 4 update. Summary: I. No. For design basis accident / malfunctions described in the USAR, the revised EOPs require operator response the same as ; that presented in the USAR. II. No . . The revised EOPs provide operator guidance should an accident / malfunction of a different type than evaluated in . the USAR occur, but do not themselves create the new possibility for such an accident / malfunction. B-1 i ,_ = . . . ._ _ - _. . t, !: ,
- p .*- +
y. III; No. The margins 7of safety are determined from the design basis accidents / malfunctions. As indicated in'I above, the' 1
- revised'EOP3 require the:same operator' response for design l basis accidents / malfunctions as-presented in the.USAR.
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