ML20137K341
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| ML20137K341 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 01/09/1986 |
| From: | Anderson C, Pullani S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20137K339 | List: |
| References | |
| 50-029-85-25, 50-29-85-25, IEIN-84-09, IEIN-84-9, NUDOCS 8601240008 | |
| Download: ML20137K341 (18) | |
See also: IR 05000029/1985025
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 50-029/85-25
Docket No. 50-029
License No. 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
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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.
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Results: No violations were identified. An item remained unresolved at the
end of inspection. For details of this item see Section 7.3.1 of this report.
8601240000 e60117 9
DR ADOCK O
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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 meeting.
2.0 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
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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
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(III.G.2.a).
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* Separation by a horizontal distance of at least 20 feet with no
intervening combustibles and with fire detection and automatic
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fire suppression installed in the fire area (III.G.2.b).
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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
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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
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concern. Any alternate or dedicated system requires NRC review and
approval prior to implementation.
For situations in which fire protection does not meet the requirements
,
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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 basis. Such exemption requests are
- submitted to the NRC for review and approval and must be justified by the
licensee on a technical basis.
4.0 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.
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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
.
.
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A sketch of the integrated SSS is shown in Attachment 2 of this
report. The NRC review of the integrated SSS is on going, and the
Safety Evaluation Report (SER) for its approval is yet to be issued.
The licensee has completed the construction of the integrated SSS and
has prepared draft procedures for its operations. However, pending
issuance of the SER, the licensee has not declared the integrated SSS
operational.
5.4 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 staff.
6.0 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 III.G.
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 plant.
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 concern. Procedures were reviewed for
general content and feasibility.
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.
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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 seals.
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 Pump.
7.0 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 conditions.
7.2 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
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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 hours.
The team did not identify any unacceptable conditions.
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
lugs.
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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 cables. . .
<
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
,
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The licensee's procedure provides timely and reliable power
for required cold shutdown equipment. s
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Based on the above, the team found the above item acceptable.
7.2.2 Procedure Walk-Through
The team walked through selected portions of the procedures to
determine that shutdown could ' be attained in an orderly and
timely fashion.
The team did not identify any unacceptable conditions.
7.3 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
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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 problems.
7.3.1 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 concern.
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-
vals.
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,
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* 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
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with 480V BUS 5.2 supply breaker,
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* 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
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* 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,
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* 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
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Battery Distribution Switchboard (BDS) 3 supply breaker to
:, Distribution Board 3A is not coordinated with the supply
breaker to BDS 3.
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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:
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'- * False motor, control, and instrument indications can occur
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such as those encountered during 1975 Brown's Ferry fire.
These could be caused by fire initiated grounds, short or
open circuits.
*
Spurious operation of : safety related or non-safety related -
components can occur that would adversely affect shutdown
capability (e.g;, RHR/RCS isolation valves).
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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
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No unacceptable conditions were identified.
7.3.3 Common Enclosure Concern
The common enclosure concern is found when redundant circuits
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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 penetrations.
A number of circuits, selected on a sampling basis, were
examined for this concern.
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 features.
No unacceptable conditions were identified.
8.0 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 thereto.
The team examined the plant emergency lighting system to ascertain the
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licensee's compliance with the above requirement.
The team did not identify any unacceptable conditions.
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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 Building.
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 operation.
Based on the above, the team found this item acceptable.
9.0 011 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
Earthquake.
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 program.
The team did not identify any unacceptable. conditions.
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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 of.
this report.
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 deviations.
An unresolved item is discussed in Section 7.3.1.
13.0 Conclusions
No violations were identified. An item regarding the breaker coordination
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study remained unresolved at the end of this inspection.
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.
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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, 1985.
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 information.
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).
4
At no time during this inspection was written material provided to the
licensee by the team.
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ATTACHMENT 1
Correspondence List
1. Letter, USNRC to All. Reactor Licensees with Plants Licensed Prior to
. January 1,1979, dated November 24,-1980.
2. 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'81-12.
3. Letter, YAEC to USNRC, FYR 81-43, dated March 19, 1981;
Subject: Compliance With Appendix R to 10 CFR 50.
4. Letter, YAEC to USNRC, FYR 81-91, dated June 15, 1981;
Subject: Additional Information on Alternative Safe Shutdown System
Proposal.
5. Letter, YAEC to USNRC, FYR 81-163, dated ' December 28, 1981;
Subject: Request for Schedule Extension.
6. 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 50.
7. Letter, YAEC to USNRC, FYR 82-50, dated May 13, 1982,
Subject: Request for Exemption.
8. Letter, YAEC to USNRC, FYR 82-75, dated July 15, 1982;
Subject: Schedule for Submittal of Additional-Informatien on Appendix R.
9. Letter, USNRC to YAEC, NYR 82-177, dated August 5, 1982;
Subject: Fire Protection - Yankee Nuclear Power Station.
10. Letter, YAEC to USNRC,.FYR 82-88, dated August 30, 1982;
Subject: Response to Request for Additional Information on
Appendix R - Alternative Safe Shutdown System.
11. Letter, YAEC to USNRC, FYR 83-40, dated April 12, 1983;
Subject: Integrated Approach to Seismic Hot Shutdown and
Appendix R Shutdown Requirements.
12. Letter, YAEC to USNRC, FYR 83-60, dated June 9, 1983;
Subject: Cooldown Capability of Proposed Integrated Shutdown Systems.
13. 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).
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14. 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
Station.
15. 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 Details.
17. Letter, YAE'.: to USNRC, FYR 84-116, dated December 28, 1984;
Subject: Appendix R Integrated System Design Details.
18. 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).
19. Letter, YAEC to USNRC, FYR 85-49, dated April 30, 1985;
Subject: Revision to Appendix R Integrated System Design Details.
20. Letter, YAEC to USNRC, FYR 85-128, dated November 7, 1985;
Subject: Revision to Appendix R Integrated System Design Details.
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D O
ATTACHMENT 2
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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 48.8
Recirculation System 10.1
Reactor Protection System and 10.0
(Includes Operator Errors)
Opera +or Failure to Manually
Depressurize MCS for Small LOCAs
if HPSI Fails 9.5
Accumulator 6.2
Operator Errors in :nitiating/
Controlling Feedwater 5.3
Failure of MCS Loop Isolation Valves
to Close During Steam Generator Tube
Rupture Plus Operator Errors
in Responding to Event 4.6
Diesel Generators Plus Steam-Driven
Emergency Feed Pump (Including Operator
Errors) - Loss of AC - 2.4
Pressurized Thermal Shock Induced Reactor
Vessel Failure Due to perator Errors
During Degradation of DC Power Events 0.5
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