ML20116G948
| ML20116G948 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 11/02/1992 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20116G932 | List: |
| References | |
| NUDOCS 9211120198 | |
| Download: ML20116G948 (22) | |
Text
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ATTACllMENT A-1 Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 203
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Revise the Technical Specification as follows:
Remove Paces Insert Paaes 3/4 9-4 3/4 9-4 B 3/4 9-1 B 3/4 9-1 7
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9211120198 921102 PDR-ADOCK 05000334 P
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y DPR-66 REFUELING 01 ERATIONS 3 /4. 9 L.CONTMNMENT BUTLDING PENETRATILNd LIMITING CONDITION FOR OPERATION m'h E
3.9 The containmc7t building penetrations shall be in the foi owing status:
s.
,$7 a.
The equisent door closed and held in place by a minimum of four bolts, b.
A minimum of one door in each a r2ock is closed, and c
Each penetration providing direct access from the containment atmesphera to the outside atmosphere shall be either:
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Closed by an isolation valve, blind flange, pf manual
,.fy valve, or Q pey4 4 e,\\,e%\\ 9%\\4 eg n
2.
Exhausting at less than or equal to 7500 cfm through
,1 OPERABLE Containment Purge and Exhaust Isolation
]
O Valves vi%d isolation times as specified in Table
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3.6-1 to Ol'ERABLE HEPA filters and charcoal adsorbers ss.,
of the Supplemental Leak Collection and Release L u em
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1 (SLCRS).
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APPLICABILITY:
During CORE ALTERATIONS or movement of irradix d
s fuel within the containment.
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With the remai rements of tile above s;h drication not satisfied, immediatel'.
wpend all operations involving CORE ALTERATIONS or mov ent ol tradiated fuel in the containment. The provisions of Sp,4CJfication 3.0.3 are not applicable.
SURVEILLANCF REQUIREMENTS 4.9.4.1 Each of the abcVe required containment penetrations shall s.e determined to be in its above, required condition within 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> prior to the start of and at least once per 7 days dt. ring CORE ALTERATIONS or movement of irradiated fuel in the containment.
w.
4.9.4.2 The containment purge and exhaust system shall be demonstrated OPERABLE by:
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a.
Verifying the flow rate through the SLCRS at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the system is in operation.
b.
Testing tha Containment Purge and Exhaust Isolation Valves per the applice.ble portions of Specification 4.6.3.1.2, a
s and c.
Testing the SLCRS per Specification 4.7.a.1.
BEAVER VALLEY - UNIT 1 3/4 9-4 Amendment No.
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"DPR-66 3/4.9 REFUELING OPERATTONS BASES J/4.9.1 BORON CONCENTRATION The limitations on minimum boron concentration (2000 ppm) ensure that:
1) the reactor will remain subcritical during CORE ALTERATIONS, and
- 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel.
The Limitation of Keff of no greater than 0.95 which includes a
conservative allowance for uncertainties, is sufficient to prevent reactor criticality during refueling Operations.
3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in tt' reactivity condition of the core.
3/4.9.3 DECAY TIME The minimum requirement for reactor suberiticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures-that sufficient time has elapsed to allow the radioactive decay of the uhort lived fission products.
This decay time is consistent with tne assumptions used in the accident analyses.
-r 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The reqmirements on containment penetration closure snd operability of the containment purge and exhaust system HEPA Ziltets and charcoal adsorbers ensure that a
release of radioactive material within containment will be restricted from leakage to the environment or filtered through the HEPA filters and charcoal adsorbers prior to discharge to the atmosphere within 10 CFR 700 limits.
The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurication potential 'hile in the REFUELING MODE.
Operations of the containment purge rnd exhaust
.=ystem HEPA filters and charcoal adsorbers and the resulting iodine removal capacity are consistent with the assumptiens of the. :ident analysis.
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Too ItuEt?T "A" t
BEAVER VALLEY - UNIT 1 B 3/4 9-1 (pphWAy
Attaenment to Containment Buildina Penetrations Insert "A"
AAl containment penetrations, except ior the containment purge and exhaust penetrations, that provide direct access from cont.ainment atmosphere to outside atmosphere must be isolated on at least one side.
Penetration closure may be achieved by an isolation valve, blind
- flange, manual
- valve, or functional equivalent.
Fur.ctional-equivalent isolation ensures releacert from the containment are prevented for credible accident scenurios.
The isolation techniques must be approved by an engineering evaluation and may include use of a
material that can provide e temporary, pressure tight seal-capable of maintaining the integrity of the penetration to restrict the relense of radioactive material from a fuel element rupture.
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k BEAVER VALLEY - UNIT 1 (Proposed Wording)
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ATTACHMENT A-2 Beaver Valley Power Station, Unit No. 2 j
Proposed Technical Specification change No. 69
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t Revise the Technical Specification as follows:
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Remove Paaes Insert Pacts 3/4 9-4 3/4 9-4 i
B 3/4 9-1 B 3/4 9-1 i
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HPF-73 REFUELING OPERATIONS CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:
a.
The equipment door closed and held in place by a minimum of four bolts, b.
A minimum of one door in each airlock is closed, and c.
Each penetration providing direct access from the containment atmosphere to the outside atmosohere snall be either:
1.
Closed by an isolation talve, blind flange, % manual valve, or 3 2.
Exhausting at less than or equal to 7500 cfe through OPERABLE Containment Purpe and Exhaust Isolation Valves with isolation times as specified in Table 3.6-1 to OPERABLE HEPA filters and charcoal adsorbers of the Supplemental Leak Collection and Release System (SLCRS).
APPLICABILITY:
During CORE ALTERATIONS or movement of irradiated fuel within the containment.
ACTION:
7 With the requirements of the above specification not satisfied, imer diately suspend all operations involving CORE ALTERATIONS or movement of irradiated fsel in the containment.
The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REOUIREMENTS 4.9.4.1 Each of the above requitec.oritainm*at penetrations shall be determined to be in its abova required condition witbi. 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> prior to the start of and at least once per 7 days during CORE At.IERATIONS or movement of irradiated fuel in the containment.
4.9.4.2 The contairwient purge and exhaust system shall be demonstrated OPERABLE by:
a.
Verifying the flow rate to the SLCRS at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the system is in operation.
b.
Testing the Containment Purge ana Exhaust Isoletion Valves per the applicable portions of Specification 4.6.3.1.2, and c.
Testing the SLCRS per Specification 4.7.8.1 with the exception of item 4.7.8.1.c.2.
BEAVER VALLEY - UNIT 2 3/4 9-4
@f'ofC 0
'NPF-73 3/4.9 REFUELI'JOPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitations on minimum boron concentration (2000 ppm) ensure that:
- 1) the reactor will remain suberitical during CORE ALTERATIONS, and 2) a uniform bcron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel.
The limitation on K,g7 of no greater than 0.95 which includes a conservative allowance for uncertainties, is sufficient to prevent reactor criticality during refueling operations.
Isolating all reactor water makeup paths from unborated water soutces pro-cludes the possibility of an uncontrolled boron dilution of the filled portions of the Reactor Coolant System.
This limitatioa is consistent with the initial conditions assumed in the accident analyses for MODE'6.
3/4.9.2-INSTRUMENTATION h
The OPERABILITY of the source range neutron flux monitors ensures that.'
redundant monitoring capability is available to detect changes in the reactivity condition of the core.
1 3/4.9.3 OECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products.
This decay time is consistent with the assumptione. used in tne accident analyses.
3/4.9.4-CONTAINMENT BUILDING PENETRATIONS s
The requirements on containment penetration closura limit leakage of radio-active material within containment to the environment tc ensure compliance with 10 CFR 100 limits.
The requirements on-operation.of the SLCRS ensure that trace-amounts of radioactive material within containment will be ffitered through HEPA filters charcoal absorbers prior to discharge to the atmosphere.
These require-ments are sufficient to restrict radioactive material release from a fuel element rupture based-upon the lack of containment pressurization potential while in the REFUELING MO0E.
(A06DSERTdi) 9 1
3/4.9.5 COMJNICATIONS The requirements for communications capability ensures that refueling station personnel can be promntly informed of significant changes in the Y
facility status or core reactivity conditions during CORE Al.TERATIONS.
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Beaver Valles - Unit 2 B 3/4 9-1 1
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4 ttachment to containment Buildina Penetrations l
Insert "B"
i All_ containment penetrations, except' for the containment purge and i
exhaust penetrations, that provide. direct access from containment i
atmosphere to outside atmosphere must be isolated on at least one
- side, Penetration _ closure may_ be achieved by an isolation valve, blind flange, ' manual
- valve, or functional equivalent.
1 nctional 2
equivalent isolation ensures releases from the contain'.:nt are j
prevented for credible accident scenarios.
The isolation techniques nust be approved by an engineering evaluation and may include _use of i
a material that can provide a_ temporary, pressure tight seal capable of maintaining the integrity of the penetration to restrict the l
release of radioactive material from a fuel element rupture.
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BEAVER VALLEI UNIT 2-(Proposed Wording) i
ATTACHMENT B Beaver Valley Power Station, Unit Nos. 1 and 2 g
Proposed Technical Specification Change No. 203 and 69 t
REVISION OF SPECIFICATION 3.9.4 TITLED
" CONTAINMENT BUILDING PENETRATIONS" s
A.
DESCRIPTION OF AMENDMENT REQUEST The proposed change would revise the Limiting condition for Ope ation (LCO) 3.9.4,
" Containment Building Penetrations," and associated Bases.
The specific revision would add the words "or a
approved functional equivalent" to LCO 3.9.4.c.1.
The Bases section for LCO 3.J.4 would be revised to add a discussion on the use of equivalent isolation methods.
B.
BACKGROUND During core alterations or moversent of irradiated fuel assembli m within containment, a
release of fission product radioactivity within the containment will be restricted from escaping to the environment, should a fuel handling accident ocriar, when the LCO 3.9.4 requirements are mct.
In Mode 6,
the potential for containment pressurization as a_
result of an accident is not likely; therefore, requirements to isolate the containment from the outside atmosphere may be different than the requirements for Modes 1
through 4.
LCO 3.9.4 requirements are referred to as
" containment closure" cather than
" containment integrity" as specified in LCO 3.6.1.1 for oporational Modes 1 through 4.
Contair. ment closure means that all potential escape paths are closed or capable of being closed.
Since the potential for containment pressurization as a
result of an accident is not
- likely, the 10 CFR Appendix J leakage criteria and tests are not required.
During a
refueling outage, various activities must be completed in the containment building.
Steam generator eddy current testing is one of these activities, and is required to be completed during each refueling outage by the plant's technical specifications.
To accomplish eddy current testing, temporary cables must be run from an area outside of containment to the steam generators located inside the containment building.
Due to the length of time required to complete eddy current testing, this activity must proceed during the periods when containment closure is required.
Therefore, a
temporary containment-penetration is installed using an existing spare electrical penetration.
The spare electrical penatration has bolted on e
removable blind flange (s) with o-ring seals.
Once the blind flange (s) are removed, an opening erists which allows passage of temporary cables into containment.
With the cables run through the spare electrical penetration, a temporarf seal is installed around the cables so that a direct access from the containment atLosphere to the outside atmosphere does not exist.
Prior to entering operational Mode 4, the spare electrical penetration is returned to its original +ondition and leak testing is performed-to ensure the requirements of 10 CFR Appendix J are met.
ATTACHMENT B, continued Proposed Technical Specification Change Nos. 203 and 69 Page 2 l
C.
JUSTIFICATION The addition of the words "or approved functional equivalent" will clarify that a
method other than the explicit use of isolation
- valves, blind flanges or manual valves are acceptable i
to ensure that containment-closure -is achieved for Mode 6
refueling activities.
An equivalent isolation of a penetration, which provides direct access from the containment atmosphere to the outside atmosphere, will ensure that any release of fission product radioactivity within the containment will be restricted from escaping to the environment.
A properly installed temporary penetration seal will provide a
containment closure during i
refueling functionally equivalent to an isolation valve, blind flange or manual valve.
An engineering evaluation will be performed on each type of temporary seal used to -meet the proposed LCO
- wording, to ensure that the seal is indeed equivalent for postulated accident scennrios during core alterations or movement of irradiated fuel in containment.-
i The proposed revisions are also consistent with wording contained in NUREG 1431
- titled,
" Standard Technical Specification For Westinghouse Plants."
Therefore, the addition of the words "or approved functional I
equivalent" will provide for the-use of temporary penetration l-seals for refueling activities only, which will be evaluated to ensure-that they will provide an acceptable method of containment closure.
The ability of the containment building to ensure that i
any release of radioactive fission prooucts will be restricted from escaping to the environment, will remain unchanged by this i
proposed amendment.
D.
SAFETY ANALYSIS The proposed revision will not adversely affect the safety of the plant.
The proposed addition of the ability to use an-equivalent I
type seal applies only during cold shutdown / refueling conditions and not while the. plant is critical.
The use of an equivalent containment penetration seal during refueling ~ operations will not create an unsafc condition or-adversely affect any-system, subsystem or component that is required to perform a safety function while in this condition.
The utilization
-o#
an engineered functionally equivalent containment penetration seal will
. provide the assurance of containment closure during refueling-activities.
The ability of the containment building to restrict the. release-of any fission product radioactivity to the environment, should a
fuel handling accident-occur, remains unchanged.
L Therefore, this-change is considered safe based on the continued.
L
' ability of the containment building to restrict the release of l
any; fission product radioactivity during ~ a
.uel hardling l
accident.
Containment: closure will be provided by an uquivalent E
containment penetration scal.
An engineering evaluation will be t
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ATTACHMENT B, contir ued Proposed Technical Specification Change Nos. 203 and 69 Page 3 performed to ensura that each type temporary seal, used in a containment penetration, will provide a. barrier which is functionally equivalent to a
blind flange,-isolation valve,- or manual valve for the conditjons that may exist during a fuel handling accident.
E.
NO SIGNIFICANT HAZARDS EVALUATION The no significant hazard considerations involved vith-the proposed amendment have been evaluated, focusing on the three standards set forth in 10 CFR 50.92(c) as quoted below:
The Commission may make a final determination, pursuant to the procedures in_ paragraph 50.91, that a proposed amendment to an operating license for a
facility licensed under paragraph 50.21(b) or paragraph 50.22 or for-a-testing facility involves no si,nificant hazards consideration,' if operation of the facility in accordance with the proposed amendment would not:
)
(1)
Involve a-significant increase-in the probability or consequences of an accident previously evaluated; or (2)
Create the possibility of a new or different kind of accident from any accident prev.ously evaluated; or -
~
(3)
Involve a significant reduction in a margin-of safety.-
The following evaluation is provided for the no_significant
'azards consideration standards.
1.
Does the change involve a
significant increase _ in the probability or consequences of-an accident previously evaluated?
-The probability of occurrence of a
previously evaluated accident is not increased beccuse failure to maintain containment closure is not an initiating condition for:a fuel handling accident.
containment penetration seal.
.-The use of an equivalent does-not ir.troduce any new potential accident initiating condition during-refueling operation.
The consequences of an accident previously evaluated is not increased because an equivalent containment penetration seal will provide -the assurance of containment closure during refueling activities.
The ability of the containment building to restrict the-release of any fission product radioactivity _to_the environment remains unchanged.
Thereforr:,
this change will not increase the probability or consequtaces of an accident previously evaluated.
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ATTACHMENT B, continued L
Proposed Technical Specification Change Nos. 203Eand 69 Page 4 l
?.
Does the change create the possibility of a new or_different kind of accident from any accident previously evaluated?
The failure of an equivalent containment penetration seal during refueling will not result in a malfunction of_any other plant equipment.
The sole purpose of establishing containment closure for refueling is to restrict the release.
of any fission product radioactivity i' the event of a fuel handling accident.
Therefcre, the prcposed changes do net create the possibility of a new or different kind of accident-from any accident previcusly evaluated.
3.
Does the change involve a significant reduction in a margin of safety?
An equivalent-containment penetration seal will provide the same assurance of containmenticlosure during refueling as a blind
- flange, isolation valve, or manual valve for credible accident scenarios.
The ability of'the containment building-to restrict the release of any fission product radioactivity i
to the environment, should a fuel handling accident occur, remains unchanged.
Therefore, the proposed change does not involve a
I significant reduction in a margin of safety.
F.
NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the considerations expressed above,_it is concluded that the-activities associated with --this-license amendment rcquest satisfies the no
'significant hazards consideration standards of 10 CFR 50.92(c)
- and, accordingly, a
no significant hazards consideration finding is justified.
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ATTACHMENT C-1 i
Beavec Valley Power Station, Unit No. 1 l
Proposed Technical Specification Change No 203 l
Typed Pages:
5/4 9-4 l
B 3/4 9-1 i
B 3/4 9-2 B 3/4 9-3 l
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DPR-66 REFUELING OPERATIONS i
3/4.9.4 CONTAINMENT BUTT.nING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrationn shall be in the following status:
i a.
The equipment door closed and held in place by a minimum of 3
four bolts, b.
A minimum of one door in each airlock is closed, and-l c.
Each penetration providing
.airect access from the j
containment atsosphere to the outside atmosphere shall be either:
1 1.
Closed by an isolation
- valve, blind flange, manual l
valve, or approved functional equivalent, or 2.
Exhausting at less than or equal to 7500 cfm through OPERABLE Containment Purge and Exhaust Isolation Valves with isolation times as specified in Table i
3.6-1 to OPERABLE HEPA filters and charcoal adsorbers i
of the Supplcmental Leak Collection and Release System j
(SLCRS).
l APPLTCABILITY:
During CORE ALTERATIONS or movement of irr&diated i
fuel within the containment.
4 ACTION:
With the requirements of the above specification not satisfied, 4
inmediately suspend all operations involving: CORE ALTERATIONS or mcVement of irradiated fuel in the containment. The provisions of
{
Specification 3.0.3 are not applicable, r
SURVEILLANCE REQUIREMENTS i
4.9.4.1 Each of the abovn rcquired containment penetrations shall be determined to be in its above required condition within ISO. hours prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment.
a 4.9.4.2 The containment
. purge and exhaust system shall be demonstrated OPERABLE by:
a.
Verifying the flow rate through the SLCRS at least cnce per; 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the system is in operation.
b.
Testing-the Containment Purge-and Exl
' Isolation Valves
^
per the applicable portions of Speca.1 cation 4.6.3.1.2, i
and a
c.
Testing the SLCRS per Specification 4.7.8.1.
BEAVER VALLEY - UNIT 1 3/49-4 Amendment ~No.
(Proposed Wording)
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DPR-66 3/4.9 REFUELIMG OPERATIONS
' BASES 3/4.9.1 BORON CONCENTRATION The limitations on minimum boron concentration (2000 ppm) ensure that:
1) t'io reactor will remain subcritical during CORE ALTERATIONS, and
- 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the raactor vessel.
The Limitation of K
f no greater-than 0.95 which includes a
conservative allowanceeff for uncertainties, is sufficient to prevent reactor criticality during refueling operations.
3/4.9.2 INSTRUMENTATlqH The OPERABILITY of the source range neutron flux monitors ensures-that redundant monitoring capability is available to detect changes in the reactivity condition of the ccre.
3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblics in the reactor vessel ensures that sufficient time has elapsed to al-low tue radioactive. decay of the shcrt lived
'ission products.
This decay time is consistent with the assumptior used in the accident analyses.
3/4.9.4 CONTAINMENT Bt1ILDING PENETRATIONS The requirements on containment penetration closure and operability B
- c. f the containment purge and exhaust system HEPT filters and-charcoal adsorbers-ensure that a
release of-radioactive material uithin containment-will be
- rebtricted from-leakage to tne environment or filtered through the HEPA-filters and charcoal adsorbers prior to discharge to the atmosphere within-10-CFR 100 limits.
The OPERABILITY and closure restrictions are sufficient to rcstrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.
Operations of the containment purge and exhaust system HEPA -filters and charcoal adsorbers and the resul. ting iodine removal capacity. are consistent with the assuinptions of: the accident analysis.
All containment penetrations, except for the <ontainment purge and exhaust penetrations, that provide direct-access from containment atmosphere to -outside l atmosphere -must be isolated'on at least one ride.
Penetration closure may be achieved by'an isolation valve, blind
- flange, manual
- valve, or functicnal equivalent.
Functional
-egaivalent isolation ensures releases from the containment are prevented _for_ credible accident _ scenarios.-- The isolation techniques must be approved'by an engineering evaluation and may include use of 3EAVER VALLEY - UNIT 1 B 3/4 9-1 Amendment No.
(Proposed Wording).
. _ _ - _ _ _ _ _ _ _ =
DPR-66 REFUELING OPERATIONS BASES
=
3/4.9.4 CONTAINMENT BUILDING PENETRATIONS (Centinued) a material that can provide a temporary, pressure tight seal capable l
of maintaining the integrity of the penetration to restrict the i
release of radicactive material from a fuel element rupture.
3/4.9.5 COMMUNICATIOJ{E The requirements for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during.
3/4.9.6 MANIPULATOR CRANE OPERABILITY The OPERABILITY requirements for the manipulator cranes ensure that:
1) manipulator cranes will be used for movement of control rods and fuel assemblies;
- 2) each crane has sufficient load capacity to lift a
control rod or fuel assembly; and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently-engaged during lifting operations.
3/4.9.7 CRAME TRAVEL - SPENT FUEL STORAGP BUILDING The restriction on movement of loads in excess of the normal weight of a-fuel assembly over other fuel assemblies ensures that no more 4
than the contents of one fuel assembly will be ruptured in the event of a fuel handling accident.
This assumption is consistent with the activity release assumed in the accident analyses.
1/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that. at least one residual heat removal (RHR) loop be in operation ensures that 1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140*F as required during the REFUELING MODE, and 2) sufficient coolant circulation is maintained throughout the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification.
The requirement to have two RHR lcops. OPERABLE When there is less than 23 feet of water above the reactor pressure vessel flange ensures that a
single failure of the operating RHR loop:will net result in a complete loss of residual' heat removal capability.
With the teactor vessel head removed and.23 feet of water above the reactor. pressure vessel flange, a large heat sink is available for core' cooling.
Thus,,in the event of a failure of the operating _RHR BEAVER ?iALLEY - UNIT 1 B 3/4 9-2 An endment No.
(Pcoposed Wording)
I m
1 DPR.
REFUELING OPERATIONS s
4 BASES i
3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANP CIRCULATION (Continued)
- loop, adequate time is provided te initiate emergency procedures to 1
cool the core.
3/4.9.9 CONTAINMENT PURGE AMD EXHAUST IlqLATION SYSTEM THE OPERABILITY of this system ensures that the containment vont and purge penetrations will be automatically isolated upon detection of high radiation leve:ls within the containment.
The integriuy of the containment penetrations of thie system is required to restrict the.
i release of radioactive material from the containment. atmosphere to acceptable levels which are less than those listed in 10 CFR 100.
Applicability in MODE 5,
although not an NRC safety requirement, will provide additional protection against small releases of a
radicactive material from the containment during maintenance activities.
3/4.9.10 AND 3/4.9.11-WATER LEVEL - REACTOR VEFT ~ AND STORAGE POOL The restrictions on minimum water level' ensure '. hat suf ficient water depth is available to remove 99%
of ths assumed 10% iodine gap activity released from the rupture of an i. radiated fuel assembly.
1 The minimum water depth is consistent w'ch the assumptions of the accident analysis.
3/4.9u12 and 3/4.9.13 FUEL BUILDING VENTILATJON SYSTEM The limitations on the storage pool ventilation system ensure that all radioactive material released from an irradiated fuel assembly i.
will be filtered through.tlua HEPA filters and charcoal acsorber i
prior to discharge to the atmosphere.
.The OPERABILITY of this system and the resulting iodine removal capacity are: consistent with the assumptions of the accident analysis.
The sport fuel pool' area ventilation system is non-safety related and only recirculates air through the fuel building.
The SLCRS portion of the ventilation i
system is safety-related and maintains a negative pressure in the fuel building.
The SLCRS flow is normally exhausted.to the atmosphere without filtering, however, the flow'is diverted through the main filter banks by manual actuation cr on a high radiation signal.
i 3/4.9.14 FUEL STORAGE - SPENT FUEh_ STORAGE POOL 4
The requirements for fuel storage in the spenr. fuel pool ensure that:
(1) the-spent fuel pool will remain subcritical during fuel storage; and (2) a uniform boron concentration 19 maintained-in the water volume in the spent fuel pool-to provide negutive reactivity far postulated accident conditions under the guidelines of ANSI BEAVER VALLEY - UNIT 1 B 3/4 9-3 Amendment No.
(Proposed Wording)
4 DPR-66 REFUELING OPERATIONS BASES 3/4.9.14 FUEL STORAGE - SPENT FUEL CTORAGE POOL (Continued) 16.1-1975.
The val;e of 0.95 or less for kaff which includes all uncertainties at the 95/95 probability / confidence level is the acceptance criteria for fuel storage in the spent fuel pool.
The Action Statement applicable to fuel storage in the spent fuel pool ensures that:
(1) the spent fuel pool is protected from distortion in the fuel storage pattern that could result in a critical array during.the movement of fuel; and (2) the boron concentration is maintained at 2 1050 ppm (this includes a 50 ppm conservative allowance for uncertainties) during all actions involving movement of fuel in the spent fue.' pool.
The surveillance Rcquireme ts applicable to fuel storage in the spent fuel pool ensure taat:
(1) the fuel assemblies satisfy the analyzed U-235 encichment limits or an analysis has been nerformed and it was determined that Keff is 5
0.95; and (2) che baron concentration meets the 1050 ppm limit.
The enrichment limitations for storage of fuel in a 3 of 4 array in the spent fuel pool is based on a nominal region average enrichment with individual fuel assembly tolerance of + or - n.05 w/o U-235.
The results of the spent fuel pool criticality analysis (August 1986) for Westinghouse STD/ Vantage SH and OFA/ Vantage 5 fuel in three of four storage locations show thal there is m re ti.an 0.3%
margin to the keff limit of 0.95 with all uncertainties included.
Based on the sensitivity study completed with this analysis, an increase in the maximum allowed enrichment for fuel stored in the spent fuel storage racks from 4.00 to 4.05 W/o will increase the maximum rack keff by less than 0.002.
Therefore, with Westinghouse 17 x
17 STD/ Vantage 5H and OFA/ Vantage 5 fuel enrici.ed at 4.05-w/o stored in the spent fuel racks in three of four storage locations and with all of the assumptions and conservatisms presented in the criticality analysis, the maximum rack keff will be less than 0.95.
3/4.9.15 CONTROL ROOM EMERGENCY-HABITABILITY SYSPEMS-
"he OPERABTLITY of the control room emergency habitability system ensures that. the control room will remain habitable for operations personnel during and. following all ciedible accident conditions.
The ambient air temperature'is controlled to prevent exceeding the-allowable equipment qualification temperature-for the equipment and instrumentation in the control room.
The OPERABILITY of this. system b
in conjunction with control room design provisions is based'on limiting the-. radiation' exposure to personnel occupying the' control room to._ 5 rem or _less whole
- body, or its equivalent.
_This limitation is consistent with the requirements ~of-General Design Criteria 19 of Appendix "A",
BEAVER VALLEY - UNIT 1 B 3/4 9-4 Amendment No.
(Proposed Wording) l
ATTACHMENT C-2 Beaver Valley Power Station, Unit'No. 2 Proposed Technical Specification Change No. 69 Typed Pages:
3/4 9-4 B 3/4 9-1 B 3/4 9-2 B 3/4 9-3 B 3/4 9-4
(
~
NP F '/ 3-REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING DENETRATIONS LIMITING CONDITION FOR OPERATION
-=-
3.9.4 The containment building penetrations shall be in the following status:
a.
The equipment door closed and held in place by a minimum of four bolts, b.
A minimum of one door in each airlock is closed, and c.
Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
1.
Closed by an isolation
- valve, blind flange, manual valve, or approved functional equivalent, or 2.
Exhausting at less than or equal to 7500 cfm through OPERADLE Containment Purce and Exhaust -Isolation Valves-with isolation times as specified in Table 3.6 to OPERABLE HEPA filt ers ac.C charcoal adsorbers of the Supplemental Leak Collection and Release Syrtem (SLCRS).
ARELI.CABILITY:
During CORE-ALTERATIONS or movement of irradiated fuel within the containment.
ACTION:
With the requirements of the above-specification not satisfied, immedictaly suspend all operations -involving CORE ALTERATIONS or movement.of irradiated fuel in the containment. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS.
4.9.4.1 Each of the above required containment penet ations shall be determined to be in its above required condition within 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in :the containment.-
4.9.4.2 The.
containment purge and exhaust. system.shall be demonstrated CPERABLE by:
a.
Verifying-the flow rate through the SLCRS at leest cnce per-24 hours when the system is in operat)cc.
b.
-Testing the Containment Purge and' Exhaust Isolation Valves per the applicable portions of Specification 4.6.3.1.2,_and c.
Testing the SLCRS per Specification 4.7.8.1 with the exception of item 4.7.8.1.c.2.
~ BEAVER VALLEY - UNIT 2 3/4 9-4
. Amendment No.
(Proposed Wording)
.__ _ _ ____ _j
i NPF */ 3-l 3/4.9 REFUELING OPERATIONS 4
BASES i
m I
3/4.9.1 BORON CONCENTRATION The limitations on minimum boron concentration (2001,pm) ensure j
that:
1) the reactor will remain _subcritici!
uring CORE i
ALTERATIONS, and
- 2) a uniform boron concentration in caint ined for reactivity control in the water volume havino direce access to the reactor vessel.
The limitation on Keff of no greater than 0.95 4
l which includes a
conservative allowance for uncertainties, is sutficient to prevent reactor-criticality during -refueling operations.
l Isolating all reactor water makeup paths from unborated water j
sources precludes the possibility of an uncontrolled boron dilution of the filled portions of the Reactor Coolant System.
This-l limitation is consistent with the' initial conditions assumed in the l
accident analyses for MODE 6.
3/4.9.2 INSTRUMENTATION f
The OPERABILITY of the source range neutron flux-monitors
)
ensures that redundant monitoring capability is available to detect-changes in the reactivity ccndition of the core.
1 l
3/4.9.3 DECAY TIlig b
l The minimum requicerent for reactor-subcriticality prior to movement of irradiated fuel assemblies ja the reactor vessel ensures j
that sufficient time has elapsed to allow the radioactive decay of the short lived fission products.
This decay time is consistent with the assumptions used in the accident analyses.
e l
3/4.9.4 CONTAINMFNT BUILDING PENETRATIONS i
The requirements on - containment penetration _ closure limit leakage of' radioactive material within containment to the i
environment to ansure compliance with 10 CPR 100 limits.
The-reqbirements on operation of the SLCRS ensure that t.. ace amounts of radioactive' material within containment will lxt filtered through HEPA-filters
. charcoal absorbers prior to-discharge to the j
atmosphere.
These requirements
.are sufficient to restrict 4
radioactive material release from a fuel element rupture based upon the
'ack of containment pressurization potential while in the REFUELING MODE.
i All containment penetrations, except-for the containment purge h
l and exhaust penetrations, that provide direct acces' from I
containment atmosphere to outside atmosphere must be isalat"d on at least one side.
Penetration closure may be achieveu 'uy an solation 1
?
BEAVER VALLEY - UNIT 2 B 3/4 9-1 Amendment No.
1 (Proposed Wording)
I
- - =. _ - -
i*
4
' ' N PF -7 3-j REFUELING.OPERAT10t@
j BASES 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS (Continued) j
- valve, blind
- flange, manual
- valve, or functional equivalent.
i Functional equivalent isolation ensures releases from the containment are prevented for credible accident scenarios.
The isolation techniques must be approved by an engineering evaluation i
and may include use of a
material that can provide a temporary, pressure tight seal capable of maintaining the integrity of the penetration to restrict the release of radioactive material from c-
- i fuel element rupture.
I j
3/4.9.5 COMMUNICATIONS i
The requirements for communications capability ensures that 4
refueling station personnel can be promptly~ informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.
)
3/4.9.6 MANIPULATOR CRANE OPERhBILITY The OPERABILITY requirements for the manipulator cranes ensure that:
1) manipulator cranes will be used for movement-of control rods and fuel assemblies; 2) each-crane has sufficient load capacity to lift a
control rod or fuel assembly; and 3) the core internals 4
and pressure vessel are protected from excessive _ lifting force in the event they are inadvertently engaged during-lifting operations, i
I 3/4.9.7 CRANE TRAVEL --SPENT FUEL STORAGE BUILDING l
The reatriction-on movement-of loads in-excess of the normal weight of a fuel assembly over other fuel' assemblies ensures that no L
more than the contents of one fuel assembly plus an additional 50 j
rods in the struck-fuel assembly will be ruptured in-the e"ent of a fuel handling -accident.
This assumption -is consistent with the activity release assumed in the accident analyses.
l 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT QIRCUL?TIOR l
The requirement that at least one residual heat-removal (RHR)
I loop be in operation ensures that 1) sufficient cooling capacity is-available to remove' decay heat and maintain the water in the reactor pressure vessel below 140'F as required during the REFUELING MODE,'
l and 2) sufficient coolant circulation is~ maintained throughout the l.
reactor core to minimize the:effect of a boron dilution incident ar.d
[
prevent boron stratification.
l l
l BEAVER VALLEY
. UNIT 2 B 3/4 9-2 Amendment No.
.(Proposed Wording)-
l
-.n-
.-,a ne-
~
NP F-7 3-EtFUELING OPERATIONS BASES
_m 3/4.9.8 RESIDUAL HEAT kEMOVAL AND COOLANT CIRCULATIQH (Continued)
The requirement to-have two RHR loops OPERABLE when there-is less than 23 feet of water above the reactor pressure vessel flange ensures that a
single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability.
With the reactor vessel nead removed and 23 feet of water above the.
reactor pressure _ vessel flange, a large heat sink is available for-
[
t core cooling.
Thus, in the event of a failure of the operating RHR
- loop, adequate time is provided to initiate emergency procedures _to cool the ccre.
3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM THE OPERABILITY of this system ensures that the centainment vent-and. purge penetrations will be automatically isolated upon detection of high radiation-levels within_the containment.
The integrity of the containment penetrations of this system
- bs. required to meet
'0 CFR 100 requirementa in the event of a fuel handling accident inside containment'.
Applicability in MODE 5, although not an NRC-safety requirement,_will provide additional protection against small ~
releases of radioactive material from the containment during maintenance activities.
3/4.9.10 AND 3/4.9.11 WATER LEVEL - REACTOR VFSSEL AND STORAGE' POOL The restrictions-on minimum water level ensure that sufficient water depth is available to remove 99%.of the, assumed 10% iodine gap activity released from -the rupture of an-irradiated fuel' assembly.
The minimum water depth is consistent with the assumptions of the accident analysis.
3/4.9.12 and 3/4.9.13 FUPL BUILDING VENTILATION SYSTEM The linitations on the storage pool ~ ventilation _ system ensure that all radioactive material released from an irradiated. fuel-assembly will be filtered-through _the HEPA filters and charcoal adsorber prior to discharge to th. atmosphere.
The OPERABILITY of this system and the resulting iodine removal capacity are consistent
~
with the assumptions of the accident analysis.
The spent fuel pool area ventilation system is non-safety related and only recirculates air through the fuel building.
The fuel building portion of the SLCRS is satety-related and continuously; filters the fuel building exhaust air.
This maintains-a negative-fpressure-in' the_ fuel building.
i BEAVER VALLEY - UNIT 2 B 3/4 9-3_
Amendment No.
(Proposed Wording) l o
NPF-73 dEFUELIJJG OPERATIONS BASES
=_
3/4.9.14 FUEL STORAGE - SPENT FUEL STORAGE POOL The requirements for fuel storage in the spent fuel pool ensure that:
(1) the sper'.
fuel-pool will remain subcritical during fuel storage; and (2) a uniform boron concentration is maintained in the water volume in the spent fuel pool to provide negative reactivity for postulated accident conditions.under the~ guidelines of ANSI 16.1-1975.
The value of 0.95 or less for K g which includes all uncertaintles at the 95/95 probability / conf! bence level is the acceptance criteria for fuel storage in the spent fuel pool, Verification that peak fuel rod burnup is less than 60 GWD/MTU is provided in the -reload evaluation-report associated _with'each fuel cycle.
The Action Statement applicable to fuel storage in the spent
'he spent fuel pool is protected from fuel prol ensures that:
(1) c distortion in the fuel storage pattern.that could result in a critical array during the mo' ament of fuel; and (2).the. boron
)
concentration is maintained at 2 1050 ppm (this includes 1a 50 ppm conservative allowance.
for.
uncertainties) during all actions involving movement of fuel in the spent fuel pool.
The Surveillance Requirements applicable to fuel. storage in the-spent fuel pool ensure that:..(1)-the fuel assemblies satisfy the analyzed U-235 enrichment limits or an analysis has been' performed and it was determined that K-is s
0.95;-
and (2) the boron concentration meets the 105u ppm-$gf1mit.
BEAVER VALLEY - UNIT 2 B 3/4 9-4 Amendment No.
(Proposed Wording)
_A