IR 05000029/1985025: Difference between revisions

From kanterella
Jump to navigation Jump to search
(StriderTol Bot insert)
 
(StriderTol Bot change)
Line 1: Line 1:
{{Adams
{{Adams
| number = ML20137K341
| number = ML20202G304
| issue date = 01/09/1986
| issue date = 04/10/1986
| title = Insp Rept 50-029/85-25 on 851216-19.No Violations Noted. Major Areas Inspected:Efforts to Comply W/Requirements of 10CFR50,App R Re Fire Protection to Achieve & Maintain Safe Shutdown in Event of Fire
| title = Ack Receipt of 860305 Ltr Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-029/85-25. Corrective & Preventive Actions Will Be Examined During Future Insp
| author name = Anderson C, Pullani S
| author name = Ebneter S
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| addressee name =  
| addressee name = Heider L
| addressee affiliation =  
| addressee affiliation = YANKEE ATOMIC ELECTRIC CO.
| docket = 05000029
| docket = 05000029
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = 50-029-85-25, 50-29-85-25, IEIN-84-09, IEIN-84-9, NUDOCS 8601240008
| document report number = NUDOCS 8604140322
| package number = ML20137K339
| document type = CORRESPONDENCE-LETTERS, NRC TO UTILITY, OUTGOING CORRESPONDENCE
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 1
| page count = 18
}}
}}


Line 19: Line 18:


=Text=
=Text=
{{#Wiki_filter:.
{{#Wiki_filter:
.
U.S. NUCLEAR REGULATORY COMMISSION
 
==REGION I==
Report N /85-25 Docket N License N DPR-03 Priority --
Category C Licensee: Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701 Facility Name: Yankee Nuclear Power Station Inspection At: Rowe, Massachusetts Inspection Conducted: Dezember 16-19, 1985 Inspector:  (/[[an  /- 9-64 S.Pullafi, tection Engineer date Also participating and contributing to the report were:
A. Coppola, Mechanical Systems Specialist, BNL K. Parkinson, Electrical Systems Specialist, BNL A. Singh, Plant Systems Branch, NRR S. Wes , lant Systems Branch, NRR Approved by: O  / 9 ffo C. J. /fderson, Chief  ' date Plant 3ystems Section, DRS
!
'
Inspection on December 16-19, 1985 (Report No-029/85-25).
 
Areas Inspected: Special, announced team inspection of the licensee's efforts to comply with the requirements of 10 CFR 50, Appendix R, Sections III.G, J. and 0, concerning fire protection features to ensure the ability !
to achieve and maintain safe shutdown in the event of a fire. The inspection involved 140 inspector hours onsite and 45 inspector hours in-office by the team consisting of 4 inspectors.
 
Results: No violations were identified. An item remained unresolved at the end of inspectio For details of this item see Section 7.3.1 of this repor e60117 9 DR ADOCK O
  .    . .
 
.
.
DETAILS 1.0 Persons Contacted 1.1 Yankee Atomic Electric Company R. Aron, Senior Operations Engineer
  *S. Flynn, I&C Engineer
  *S. Fournier, Lead Systems Engineer W. Joes, Engineering Manager
  *T. Henderson, Technical Director
  *J. Kay, Technical Services Manager
  *J. Lynch, Engineer (Systems)
  *G. Papanic, Licensing Engineer
  *J. Piela, Engineer (Electrical)
E. Sawyer, Manager, Engineering Services
  *N. St. Laurent, Plant Superintendent 1.2 Nuclear Regulatory Commission (NRC)
1  J. Clifford, Licensing Project Manager, NRR
! *H. Eichenholz, Senior Resident Inspector
'
J. Stang, NRR
  * Denotes those present at the exit meetin .0 Purpose This inspection was to ascertain that the licensee is in conformance with 10 CFR 50, Appendix R, Sections III.G, J, and 0, including exemptions approved by the Office of Nuclear Reactor Regulation (NRR).
 
3.0 Background 10 CFR 50.48 and .10 CFR 50, Appendix R became effective on February 17, 1981. Section III.G of Appendix R requires that fire
;'
protection be provided to ensure that one train of equipment necessary to achieve and mair.tain safe shutdown remains available in the event
;
of a fire at any location within a licensed operating facility. For hot shutdown conditions, one train of the systems necessary must be free of fire damage (III.G.I.a). For cold shutdown conditions, repair is
,. allowed using in place procedures and materials available onsite with
;
the provision that cold shutdown be achievable within 72 hours of the initiating event (III.G.1.b). Section III.G.2 lists specific options as i follows to provide adequate protection for redundant trains of equipment located outside of the primary containment:
) *
Separation by a fire barrier having a three hour rating
 
  (III.G.2.a).
 
I-
 
  .. .. . . _ - .-
.
.
 
* Separation by a horizontal distance of at least 20 feet with no intervening combustibles and with fire detection and automatic
.
fire suppression installed in the fire area (III.G.2.b).
 
i *
Enclosure of one train in a fire barrier having a one hour rating in addition to having fire detection and automatic suppression installed in the fire area (III.G.2.c).
 
For non-inerted primary containment, Section III.G.2 specifies one of the above three protection options, or any of the following:
* Separation by a horizontal distance of at least 20 feet with no
'
intervening combustibles or fire hazards (III.G.2.d).
 
*
Fire detection and automatic fire suppression installed in the fire area (III.G.2.e).
 
*
Separation of redundant trains by a non-combustible radiant energy shield (III.G.2.f).
 
If the protection required by Section III.G.2 is not provided or the systems of concern are subject to damage from fire suppression activities, Section III.G.3 of the rule requires that an alternate or dedicated shutdown capability be provided which is independent of the area of
 
concern. Any alternate or dedicated system requires NRC review and approval prior to implementatio For situations in which fire protection does not meet the requirements
,
'
of Section III.G, however, such protection is deemed to be adequate by the licensee for the specific situation, the rule allows the. licensee to request an exemption on a case-by-case basi Such exemption requests are
; submitted to the NRC for review and approval and must be justified by the licensee on a technical basi .0 Correspondence
'
All correspondence between the licensee and the NRC concerning compliance with Sections III.G, J and 0 was reviewed by the inspection team in pre-paration for the site visit. Attachment I to this report is a listing of the correspondence reviewed.
 
f '
i
-- -
  - - , . . . . . , . . , - , -. _ - - ~ ~
 
.
.
 
5.0 Post-Fire Safe Shutdown Capability 5.1 Systems Required for Safe Shutdown The following systems are used to provide shutdown capability with a loss of off-site power:
Hot Shutdown:
* Emergency Feedwater System (EFWS)
*
Emergency Atmospheric Steam Dump System (EASDS)
*
Pressure Control and Relief System (PCRS)
* Charging and Chemical Shutdown System (CCSS)
* Primary Instrumentation
* Emergency Power System (EPS)
Cold Shutdown:
*
Shutdown Cooling System (SCS)
*
Component Cooling system (CCS)
* Service Water System (SWS)
*
Emergency Power System (EPS) (same as hot shutdown)
5.2 Fire Areas Required for Safe Shutdown The Yankee Plant has been broken down into the following nine fire areas, each of which contains safe shutdown equipment:
* Vapor Container
*
Primary Auxiliary Building
*
Diesel Generator Building
*
Nonreturn Valve Enclosure
* Screenwell Pump House
*
Main Control Room a Switchgear Room
* Cable Spreading Room
*
Turbine Building 5.3 Areas Where Alternate Safe Shutdown is Required For the following six of the above nine fire areas, the licensee proposed an integrated and dedicated safe shutdown system (SSS) to meet the Alternate Safe Shutdown requirements of Appendix R, Section III.G.3, as well as seismic hot shutcown needs under the Systematic Evaluation Program (SEP):
*
Diesel Generator Building
*
Main Control Room    -
* Switchgear Room
*
Cable Spreading Room
*
Turbine Building
*
Primary Auxiliary Building
 
.
.
 
A sketch of the integrated SSS is shown in Attachment 2 of this repor The NRC review of the integrated SSS is on going, and the Safety Evaluation Report (SER) for its approval is yet to be issue The licensee has completed the construction of the integrated SSS and has prepared draft procedures for its operation However, pending issuance of the SER, the licensee has not declared the integrated SSS operationa .4 Remaining Plant Areas All other safe shutdown areas of the plant are required to comply with Section III.G.2 of Appendix R, unless an exemption request has been approved by the staf .0 Inspection Methodology The inspection team examined the licensee's capabilities for separating and protecting equipment, cabling and associated circuits necessary to achieve and maintain hot and cold shutdown conditions. This inspection sampled selected fire areas which the licensee had identified as being in compliance with Section II The following functional requirements were reviewed for achieving and maintaining hot and cold shutdown:
* Reactivity control
*
Pressure control
*
Reactor coolant makeup
* Decay heat removal
* Support systems
*
Pro;ess monitoring The inspection team examined the licensee's capability to achieve and maintain hot shutdown and the capability to bring the plant to cold shut-down conditions in the event of a fire in various areas of the plan The examination included a review of several drawings, safe shutdown pro-cedures and other documents. Drawings were reviewed to verify electrical independence from the fire areas of concer Procedures were reviewed for general content and feasibilit Also inspected were fire detection and suppression systems and the degree of physical separation between redundant trains of Safe Shutdown Systems (SSSs). The team review included an evaluation of the susceptibility of the SSSs for damagt from fire suppression activities or from the rupture
, or inadvertent operation of fire suppression systems.
 
i
 
.
.
 
The inspection team examined the licensee's fire protection features pro-vided to maintain one train of equipment needed for safe shutdown free of fire damage. Included in the scope of this effort were fire area bounda-ries, including walls, floors and ceilings, and fire protection of open-ings such as fire doors, fire dampers, and penetration seal The inspection team also examined the licensee's compliance with Section III.J, Emergency Lighting, and Section III.0, Oil Collection System for Reactor Coolant Pum .0 Inspection of Protection Provided to Safe Shutdown Systems 7.1. Protection in Various Fire Areas The team reviewed the protection provided for selected safe shutdown system and components in the following fire areas, for compliance with Appendix R, Section III.G.1, 2, and 3,
* Vapor Container
*
Primary Auxiliary Building
*
Diesel Generator Building
*
Nonreturn Valve Enclosure
*
Screenwell Pump House
*
Main Control Room
. Switchgear Room
*
Cable Spreading Room a
Turbine Building The safe shutdown systems and components selected for inspection in the above fire areas were based on their relative importance to safety using Probabilistic Risk Assessment (PRA) techniques (see Section 11.0 for details).
 
The team did not identify any unacceptable condition .2 Safe Shutdown Procedures 7.2.1 Procedure Review The team reviewed the following safety shutdown procedure:
  *
OP-3017, Fire Emergency, Revision 8
  *
OP-3018A, Plant Shutdown with the SSS, Revision 7
  *
OP-5801, Cable Replacement for Appendix R Equipment In Order to Achieve Cold Shutdown Equipment within 72 Hours, Revision 0
 
.
.
 
The scope of this review was to ascertain that the shutdown could be attained in a safe and orderly manner, to determine the level of difficulty involved in operating equipment, and to verify that there was no dependence on repairs for achieving hot shutdown. For purpose of the review, a repair would include installing electrical or pneumatic jumpers, wires or fuses to perform an action required for hot shutdown. For cold shutdown, repair is allowed using in place procedures and materials available onsite with the provisions that cold shutdown be achievable within 72 hour The team did not identify any unacceptable condition Cable Replacement to Achieve Cold Shutdown Within 72 Hours IE Information Notice 84-09, lessons learned from NRC Inspections of fire protection safe shutdown systems (10 CFR 50, Appendix R), dated February 13, 1984, paragraph XI, fire protection features for cold shutdown systems states:
" Procedures for repairing damaged equipment should be prepared in advance with replacement equipment (i.e.,
cables made up with terminal lugs attached) stored on site in a controlled manner. All repairs should be of sufficient quality to assure safe operation until the plant is restored to an operating condition. Repairs not permitted include the use of clip leads in control panels (which means that hard-wired terminal lugs must be used),
and the use of jumper cables other than those fastened with terminal lugs."
 
YAEC Procedure OP-5801, Cable Replacement for Appendix R Equipment In Order to Achieve Cold Shutdown Within 72 Hours, Revision 0, provides procedures for installing jumper cables for the following cold shutdown components:
*
P19, Shutdown Cooling Pump
=
P23, LPST Cooling Pump
=
P20-1, Component Cooling Pump No. 1
*
P20-2, Component Cooling Pump No. 2
*
P6-1, Service Water Pump No. 1
*
P6-2, Service Water Pump No. 2
*
P6-3, Service Water Pump No. 3 The procedure OP-5801 requires installing the jumper cables by splicing one end of the jumper cables to existing cables and installing the other end to motor connectors using terminal lug c
.    ,
    ,.
.
 
L Physical in plant inspection of the stored jumper cables veri '
s fied that cable splicing kits and terminal lug kits were stored with the pre positioned jumper cables. The licensee has not installed terminal lugs on the cables. The' licensee has experienced corrosion problems with exposed terminal lugs and has determined that installing the terminal lugs at the time of jumper installation will require less time than correcting termi-nal lug corrosion problems'. To minimize (be Jength of jumper cables, the licensee has planned to splice ode ena of the jumper cable to existing cables vice using terminai'7ags on both ends ,
of the jumper cable .  .
      <
Although the licensee's procedure for cable replacement to achieve cold shutdown within 72 hours is'at variance with the guidance provided in IE Information Notice 84-09, it is con-sidered acceptable for the following reasons:
J
*
Jumper cable installation for a given ' fire will be required for no more than one pump,
*
The licensee can install required jumper cabi s in four to six hours by existing procedures, i.e., before the cold shutdown components are required for safe shutdown, and
      ,
*
The licensee's procedure provides timely and reliable power for required cold shutdown equipmen s
      >
Based on the above, the team found the above item acceptabl .2.2 Procedure Walk-Through The team walked through selected portions of the procedures to determine that shutdown could ' be attained in an orderly and timely fashio The team did not identify any unacceptable condition .3 Protection for Associated Circuits Appendix R, Section III.G, requires that protection be provided for
,- associated circuits that could prevent operation or cause malopera-tion of redundant trains of systems necessary for safe shutdown. -The circuits of concern are generally associated with safe shutdown circuits in one of three ways:
      ,
a Common bus concern
*
Spurious signals concern
* Commor, enclosure concern
 
.
.
 
The associated circuits were evaluated by the team for common bus, spurious signals, and common enclosure concerns. Power, control, and instrumentation circuits were examined for potential problem .3.1 Common Bus Concern The common bus concern is found in circuits, either safety related or non-safety related, where there is a common power source with shutdown equipment and the power source is not electrically protected from the circuit of concer The team examined, on a sampling basis, 2400V, 480V, 120V AC and 125V DC bus protective relay co-ordination. The team also examined, on a sampling basis, the protection for specific instrumentation, controls, and power circuits, including the coordination of fuses and circuit breakers. The licensee has been performing ralay settings at approximately 18 month inter-val No unacceptable conditions were identified except as follows:
Breaker Coordination Study The licensee has performed a breaker coordination study as part of the associated circuit analysis. The breaker coordination study identified the following coordination problems:
*
EMCC-1 breaker for Transformer A is not coordinated with EMCC-1 alternate feed breaker,
*
480V EBUS 2 breaker for LPSI punp 2 is not coordinated with 480V EBUS2 supply breaker,
*
480V BUS 4.1 supply breaker for MCC4-BUS 2 is not coordi-nated 480V Bus 4.1 supply from SSTR4,
*
EMCC6 supply breaker to vital bus distribution panel is not coordinated with EMCC6 supply breaker,
*
EMCC4 supply breaker to EMCC6 is not coordinated with EMCC4 supply breaker,
*
480V EBUS 3 supply breaker for LPSI pump 3 is not coordi-nated with 480V EBUS3 supply breaker, a
MCC4 BUS 1 supply breaker for the yard area fuel pit crane is not coordinated with MCC4 BUS 1 supply breaker,
*
480V BUS 5.2 supply breaker for MCC4 BUS 1 is'not coordi-nated with 480V BUS 5.2 supply breaker,
    ,
  --
 
      .
e
'
i )
,.
. ,s
  * MCC] FUS 1. supply breaker for Power P.ecepticle VC is not coordi,nated with MCC1 BUS 1 supply breaker,
  * 480V BUS 5.2 supply break'er to MCC1 BUS 1 is not coordinated
  ,
  -
with 480V BUS 5.2 supply breaker, s
  * Distribution Panel 3B supply to 480V Test Panel is not coordinated with the supply to Distribution Panel 3B,
  * EMCC3 supply breaker to EMCC5 is not coordinated with the supply breaker to EMCC3,
  *
480V EBUS 1 supply breakerg t o LPSI pump 1 is not coordinated with the supply breakers to 480V EBUS 1,
  *  supply breaker to vapor container crane CR-2 MCC1 BUS 2;dinatedwithMCC1 BUS 2supplybreaker, is not coor
 
  * 480V BUS 6.3 supply breaker to MCC1 BUS 2 is not coordinated with the supply breaker to 480V BUS 6.3,
  * MCC2 BUS 1 suppTy breaker to the Information Center is not coordinated with the supply breaker to MCC2 BUS 1, ff
''  _
  * 480V BUS 6.3' supply breaker to MCC 2 BUS.1 is not j  coordinated with the supply breaker to 480V BUS 6.3, and
  '
e *
Battery Distribution Switchboard (BDS) 3 supply breaker to
  :,  Distribution Board 3A is not coordinated with the supply breaker to BDS '-
The licensee is evaluating the results of the breaker coordi-nation study to develop a corrective action plan to resolve
  *
breaker coordination problems. The licensee committed to com-plete the evaluation within 90 days from December 20, 1985, and to inform the NRC of their corrective action plan. This is an f  unresolved item pending completion of the licensee action and its review by,NRC (85-029/85-25-01).
 
7.3.2 Spurious Signals Concern The spurious signals concern is made up of 2 items:
    ~
'
  '- * False motor, control, and instrument indications can occur
.
"
such as those encountered during 1975 Brown's Ferry fir These could be caused by fire initiated grounds, short or open circuit *
Spurious operation of : safety related or non-safety related -
components can occur that would adversely affect shutdown capability (e.g;, RHR/RCS isolation valves).
 
'
    ; ,
    ~
M/_    'M, _ ,  ,
 
. -
.
.
 
The team examined, on a sampling basis, the following areas to ascertain that no spurious signals concern exists:
  * Current transformer secondaries
  * High/ low pressure interface General fire instigated spurious signals
'
  *
No unacceptable conditions were identifie .3.3 Common Enclosure Concern The common enclosure concern is found when redundant circuits
'
are routed together in a raceway or enclosure and they are not electrically protected or when fire can destroy both circuits due to inadequate fire barrier penetration A number of circuits, selected on a sampling basis, were examined for this concer No unacceptable conditions were identified.
 
;
7.4 General Fire Protection Features The team examined the general fire protection features in the plant provided to maintain one train of safe shutdown equipment free of fire damage. Included in the scope of this effort were fire area boundaries, including walls, floors and ceilings, and fire protection of openings such as fire doors, fire dampers, and penetration seals, fire protection systems, and other fire p-otection feature No unacceptable conditions were identifie .0 Emergency Lighting 10 CFR 50, Appendix R, Section III.J, requires that emergency lighting units with at least an 8-hour battery power supply shall oe provided in all areas needed for operation of safe shutdown equipment and in access and egress routes theret The team examined the plant emergency lighting system to ascertain the
'
licensee's compliance with the above requiremen The team did not identify any unacceptable conditions.
 
    -  ~ 4 e
 
.
i
      ;
-
      \
 
Emergency Lighting in Yard Areas The team noted that the emergency lighting with 8 hour battery power supply was not provided in the yard areas between the Turbine Building, Switch Yard, and the new Integrated Safe Shutdown System Buildin However, the licensee provided adequate technical justification for not providing the required emergency lighting in those areas. The yard areas have been provided with sufficient normal lighting. In the event of a loss of offsite power, the emergency lighting in the yard areas will be provided by the security diesel generator which is located in a separate outside building and independent of the plant fire areas of concern and requiring the use of the yard areas.
 
. In addition, the licensee has plans to store hand held battery operated flash light units in the control room along with other apparatus required during safe shutdown operatio Based on the above, the team found this item acceptabl .0 011 Collection System for Reactor Coolant Pump 10 CFR 50, Appendix R, Section III.0, requires that the reactor coolant pump shall be equipped with an oil collection system if the containment is not inerted during normal operation. The oil collection system shall be so designed, engineered, and installed that failure will not lead to fire during normal or design basis accident conditions and that there is reasonable assurance that the system will withstand the Safe Shutdown Earthquak The main coolant circulating pumps in this plant are of the canned-rotor type. Each pump unit consists of a hermetically sealed motor and centri-fugal pump impeller mounted on a single shaft as an integral unit complete with heat exchanger, volute, and high pressure motor _ terminals. The pumps are designed for single-speed operation and are water lubricated. Because .
the pumps are water lubricated, there is no lube oil associated with the pumps. The requirements of Appendix R, Section III.0, Oil Collection System for Reactor Coolant Pumps are, therefore, not applicable to this plant.
 
, 10.0 Quality Assurance During the course of the inspection, the team reviewed several drawings, fire hazard analysis, fire protection modification packages, procedures, and other fire protection documents. The scope of review included verification of their technical adequacy, appropriate reviews, design and procurement controls, and other Quality Assurance requirements for the licensee's fire protection progra The team did not identify any unacceptable. condition , .
.
 
,
11.0 Use of PRA Techniques for the Inspection The safe shutdown systems (SSSs) and components were selected for inspec-tion of their fire protection features, based on their relative importance to safety using PRA techniques as opposed to random sampling basis (see Section 7.1 of this report). If there exists a plant specific PRA or related studies such as Interim Reliability Evaluation Program (IREP) or Reactor Safety Study Methodology App;ication Program (RSSMAP), their results can be used in the selection process. In the-absence of these
, types of plant specific studies generic conclusions on the relative importance of PWR and BWR systems based on a study of 15 published PRAs
!
can be used for this purpose, with lesser accuracy (see NUREG-1050, PRA Reference Document, September 1984, Section B.3.2, Figure B-5).
 
A plant specific PRA study for the Yankee Plant had been performed by the licensee and was published in a document entitled Yankee Nuclear Power Station Probabilistic Safety study, dated March 2, 1983. The relative importance of the plant systems on the basis of their percentage contri-bution to core melt frequency for all accident sequences studied is shown in Table 8.4 of the above document which is reproduced as Attachment 3 o this repor Comparing the systems in Attachment 3 to their functionally equivalent SSSs listed in Section 5.1, the following SSSs were selected for this inspection, based on their relative importance:
*
Charging and Chemical Shutdown System (CCSS)
*
Pressure Control and Relief System (PCRS)
*
Emergency Feedwater System (EFWS)
*
Emergency Power System (EPS)
In addition, some of the other SSSs-were also inspected because of their importance in the case of a fire scenario.
 
.
12.0 Unresolved Items Unresolved items are matters for which more information is -required in order to ascertain whether they are acceptable, violations, or deviation An unresolved item is discussed in Section 7. .0 Conclusions No violations were identified. An item regarding the breaker coordination
>
study remained unresolved at the end of this inspectio The details of the unresolved item are discussed-in Section 7.3.1 of this report. The licensee actions and commitments, including actions required
,
to resolve the unresolved item, are discussed in that1section.
 
i
    - - . . .
 
    .
~
.l
.
 
I'
14.0 Exit Interview The inspection team met with the licensee representatives, denoted in Paragraph 1, at the conclusion of the inspection on December 19, 198 The team leader summarized the scope and findings of the inspection at
: that time.
 
d The team leader and the licensee discussed the contents of this inspection report to ascertain that it did not contain any proprietary informatio The licensee agreed that the inspection report may be placed in the Public Document Room without prior licensee review for proprietary.information (10 CFR 2.790).
 
At no time during this inspection was written material provided to the licensee by the team.
 
,
f
$
 
. .  . - _ - _ _ _ _ .- , .. . . . _ . . . . . .- . ~ . - .. - . . . . _____
.
.
ATTACHMENT 1 Correspondence List Letter, USNRC to All. Reactor Licensees with Plants Licensed Prior to
. January 1,1979, dated November 24,-198 . Letter, USNRC to All Power Reactor Licensees with Plants Licensed Prior to January 1,1979, dated ' February 20, 1981; Subject: Fire Protection Rule - Generic Letter'81-1 . Letter, YAEC to USNRC, FYR 81-43, dated March 19, 1981; Subject: Compliance With Appendix R to 10 CFR 5 . Letter, YAEC to USNRC, FYR 81-91, dated June 15, 1981; Subject: Additional Information on Alternative Safe Shutdown System Proposa . Letter, YAEC to USNRC, FYR 81-163, dated ' December 28, 1981; Subject: Request for Schedule Extensio . Letter, USNRC to YAEC, NYR 82-92, dated April 28,- 1982; Subject: Fire Protection Rule - 10 CFR 50.48(c)(5) - Alternative-Safe Shutdown - Section III.G.3 of Appendix R to 10 CFR 5 . Letter, YAEC to USNRC, FYR 82-50, dated May 13, 1982, Subject: Request for Exemptio . Letter, YAEC to USNRC, FYR 82-75, dated July 15, 1982; Subject: Schedule for Submittal of Additional-Informatien on Appendix . Letter, USNRC to YAEC, NYR 82-177, dated August 5, 1982; Subject: Fire Protection - Yankee Nuclear Power Statio . Letter, YAEC to USNRC,.FYR 82-88, dated August 30, 1982; Subject: Response to Request for Additional Information on Appendix R - Alternative Safe Shutdown Syste . Letter, YAEC to USNRC, FYR 83-40, dated April 12, 1983; Subject: Integrated Approach to Seismic Hot Shutdown and Appendix R Shutdown Requirement . Letter, YAEC to USNRC, FYR 83-60, dated June 9, 1983; Subject: Cooldown Capability of Proposed Integrated Shutdown System . Letter, USNRC to All Licensees and Applicants.of Nuclear Power Reactors, NYR 83-200, dated October 19,'1983; Subject: .NRC Positions on Certain Requirements of Appendix R to 10 CFR 50 (Generic Letter 83-33).
 
.
.
 
1 Letter, USNRC to YAEC, NYR 83-220, dated November 14, 1983; Subject: Proposed Integration of SEP Hot Shutdown System and Appendix R-Alternate Safe Shutdown System Requirem2nts - Yankee Nuclear Power Statio . IE Information Notice No. 84-09, dated February 13, 1984, Lessons Learned from NRC Inspections of Fire Protection Safe Shutdown Systems (10 CFR 50, Appendix R).
 
16. Letter, YAEC to USNRC, FYR 84-20, dated February 15, 1984; Subject: Appendix R Integrated System Design Detail . Letter, YAE'.: to USNRC, FYR 84-116, dated December 28, 1984; Subject: Appendix R Integrated System Design Detail . Letter, USNRC to All Power Reactor Licensees and All Applicants for Power Reactor Licenses, NYR 85-16, dated January 9.1985; Subject: Fire Protection Policy Steering Committee Report (Generic Letter 85-01).
 
1 Letter, YAEC to USNRC, FYR 85-49, dated April 30, 1985; Subject: Revision to Appendix R Integrated System Design Detail . Letter, YAEC to USNRC, FYR 85-128, dated November 7, 1985; Subject: Revision to Appendix R Integrated System Design Detail I
 
D O
ATTACHMENT 2 ss-sv-too TK-58 W
      ,
s s-v=2 F o    w NI    ,.,ss-v-ses aso ML  sig [r E n Fr JL O
a g" ,
        ,ss-v as3
      " *
ss-v-4to  o .-N- n r 'r'
 
      {; .
      * a :
cs-v-4oo E Im E I
P-82  m a 7'
      * * 4
        *
7  8  CH- W-773
      '
        *
Y'k-P uls a tion  a 5I
_  Dampenor as-v-aso ss-v-2so 1r    F)4  fh e
CH AR GING LINE JL    ss-v. ass  o
*
  ;j; ss-y-se4 T,g,,y,,,,
  ,
e  s-V-aso A FI Jg
    "
I    c - ss-v-asa ass JL  '  ,
TK-55  >
a O ; : : -H- -
a  To Ulowdown 3 50,00 0 G A L _ pg  *
lleador 4  7 ss-v 1f ss-v  XT    vo-v-issi ass JL 1so T: &
    *
      &
LA F'
w f-
_
T1 X  ss-v-sto    i 7 ss-v-too P-81  Ss-y-sis
        \ I A L La S AFE GHUTOOWN SYSTEM
        -
          <
i l
          .
 
.:
.
ATTACHMENT 3-(Reproduced from the YNPS Probabilistic Safety Study, Table 8-4)
System Contributions to Core Melt Frequency Contribution to Core-System (or Action)  Melt Frequency (1%)
HPSI and LPSI Systems  4 Recirculation System  1 Reactor Protection System and  1 (Includes Operator Errors)
Opera +or Failure to Manually Depressurize MCS for Small LOCAs if HPSI Fails Accumulator Operator Errors in :nitiating/
Controlling Feedwater Failure of MCS Loop Isolation Valves to Close During Steam Generator Tube Rupture Plus Operator Errors in Responding to Event Diesel Generators Plus Steam-Driven Emergency Feed Pump (Including Operator Errors) - Loss of AC - Pressurized Thermal Shock Induced Reactor Vessel Failure Due to perator Errors During Degradation of DC Power Events  0.5
}}
}}

Revision as of 00:29, 15 December 2020

Ack Receipt of 860305 Ltr Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-029/85-25. Corrective & Preventive Actions Will Be Examined During Future Insp
ML20202G304
Person / Time
Site: Yankee Rowe
Issue date: 04/10/1986
From: Ebneter S
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To: Heider L
YANKEE ATOMIC ELECTRIC CO.
References
NUDOCS 8604140322
Download: ML20202G304 (1)
v • d • e


Text

Retrieved from "https://kanterella.com/w/index.php?title=IR_05000029/1985025&oldid=3036235"