ML20043B688
| ML20043B688 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 05/23/1990 |
| From: | Mcdonald D Office of Nuclear Reactor Regulation |
| To: | Creel G BALTIMORE GAS & ELECTRIC CO. |
| References | |
| TAC-76470, TAC-76741, NUDOCS 9005310133 | |
| Download: ML20043B688 (9) | |
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.May 23, 1990 p;
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. Mr. G. C. Creel 1
'Vice President - Nuclear Energy
~ Baltimore Gas and Electric Company
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Calvert Cliffs Nuclear Power Plant
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'MD Rts 2 & 41 i
s P. 0.: Box 1535
. Lusby,. Maryland 20657 i
Dear Mr. Creel:
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SUBJECT:
- C.'iANGES'TO THE CALVEDT CLIFFS NUCLEAR POWER PLANT, UNITS 1 AND 2,
' TECHNICAL SPECIFICATION BASES, TAC NOS 76470 (UNIT 1) AND 76741 (UNIT 2)'
By letter dated September 29, 1989, you recuested changes to the Bases a
r Sections of the'Calvert Cliffs, Units 1 anc 2,-Technical Specifications.
l fourl changes requested:and justification'for the changes are:
. The' 2
s Change No.'15.Pages-B 3/4 6-2, Operating. License Nos. OPR-53 and DPR-69 V
! AnLadministrative discrepancy was identified between the Updated Final Safety 1
j
. Analysis Report '(UFSAR) Chapter.14.20.4'.2, Re-evaluation of LOCA, and Technical
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Specification Bases 3/4'.6.1.4,~ Internal Pressure. Your further evaluation 1
. determined clarification..to both the UFSAR and: the Technical Specification.
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Bases.is-required. The Bases change will clarify that the maximum calculated i
containment pressure for a primary system break'is 47.6 psig, assuming.an
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l initial containment pressure of 14.7 psia. The limit of 1.8.psig for initial
- {
h positive containment; pressure will limit the' total pressure to 49.4-psig, iIn-additionL the Bases change will: state that the maximum-calculated containment
. pressure of a steam line break is 49.2 psig assuming an initial containment
. pressure of 16.5 psia--(1.8 psig).
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Change No.L2 - Pages B-3/4 6-3, Operating License Nos. DPR-53 and DPR 69 j
i
- Technical-Specification 3.6.1.7, Containment Purge Supply and Exhaust-I h'
Isolation Valves, does not have a supporting Bases.
The new Bases will state i
L that this Technical. Specification was developed to address a concern that theseL T
. valves' might not be capable of closing during LOCA conditions and, therefore; will be maini.ained shut by isolating air to their air operators and-3
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maintaining their air supply solenoids de-energized.
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n Change No.~3 - Pages B 3/4 6-3, Operating License Nos. DPR-53 and DPR-59
.The existing Technical specification Bases 3/4.6.2.2 implies the containment
. cooling system must be operated simultaneously with.the containment spray system in order to ensure adequate heat removal capacity is available during 1
,y 4 post-LOCA conditions. The Bases change will clarify that the containment j
cooling system will ensure that adequate heat removal capacity is available e
,when the operability requirements of the Technical Specification are met, j
. Change No. 4 - Page B 3/4-2-2, Operating License No. DPR-53 Technical Specification Bases 2.1.1 does not reflect the use of the Extended
. Statistical Combination of Uncertainties (ESCU) methodology. The ESCU methodology was used in the Unit 1, Cycle 10, analyses to demonstrate additional thermal margin to the departure from nucleate boiling (DNB) Specified AcceptableFuelDesignLimit(SAFDL). Additionally, this change corrects the j
'value of the DNB heat flux ratio (DNBR) from 1.21 to 1.15.
This Bases page i
should have been revised by the Unit 1, Cycle 10, submittal, j
A. Med in your request, these changes correct discrepancies in the existing Bases Section.. The changes reflect the existing design basis for the facility e umented in the UFSAR and design modifications previously approved by the M t tsff, therefore; we _ht _ o objections to the-proposed Technical i
Spufication Bases changes discussed above.
Enclosed are an instruction. sheet and revised Bases pages for Operating License Nos. DPR-53 and DPR-69. This concludes our action relating to the reference TAC numbers.
Sincerely, ii ORIGINAL SIGNED bye i
Daniel G. Mcdonald, Senior Project Manager
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Project Directorate I-1 Division of Reactor Projects - III Office of Nuclear Reactor Regulation
Enclosures:
As stated cc w/ enclosures: See next page Distrib1 tion:s BDectet He t.1 RAr' 7ra ACRS(10)
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DOCUMENT NAME: MEMO TO CREEL TAC 76470/76741 W
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Mr. 4. C. Creel Baltimore Gas & Electric Company Calvert Cliffs ".
.- awer plant 1
CC:
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Mr. William T. Bowen, Dresident
-Calvert County Board of Mr. Joseph H. Walter Commissioners-Engineering Division--
l Dublic Service Comission of Maryland Drince Frederick, Maryland 20678 American Building 231 E. Baltimore Street
- 0. A. Brune, Esq.
Baltimore, Maryland 21202-3486 General-Counsel Baltimore Gas and Electric-Company Ms. Kirsten A. Burger. Esq.
P. O. Box 1475 Maryland-People's Counsel _-
Baltimore, Maryland 21203 American Building, 9th Floor Mr.' Jay E.'Silberg, Esq.
231 E. Baltimore Street Baltimore, Maryland 21202 Shaw, Pittman, Potts and Trowbridge 2300 N Street,.NW Ms. Patricia ~Birnie Washington, DC. 20037 Co-Director Maryland Safe Energy Coalition-P. O. Box 902 Ms. G. L. Adams, licensing Columbia, Maryland 21044 Calvert Cliffs Nuclear-Power Plant HD Rts 2 & 4 P. O. Box 1535 1
Lusby, Maryland 20657
_l Resident !nspector i
- c/o ~ ll.S. Nuclear Regulatory Comission
.P. O. Pox 437 i
Lusby, Maryland 20657 Mr. Richard McLean Administrator - Radioecology Department of Natural Resources 580-Taylor _ Avenue Tawes State Office Building PPER B3 Annapolis,- Maryland 21401 Pegional Administrator, Region I U.S. Nuclear Regulatory Comission 475 Allendale Road-King of Prussia, Pennsylvania 19406 l
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FACILITY OPERATING LICENSE NOS. OPR-53 AND DPR
- DOCKET NOS. 50-317~AND 50-317
- Revise Appendix A'as follows:
Remove Pages Insert Pages
~B 3/4 6-2 (DPR-53)
B 3/4 6'-2 03/46-2.(DPR-69)-
B 3/4 6-2 B 3/4 6-3 (DPR-53)
B 3/4 6-3 8 3/4 6-3 (DPR-69)'
B 3/4 6-3 5 3/4 2-2 (DPR-53)
B 3/4 2-2 f-
e.
- A
. CONTAINMENT-SYSTEMS-9i BASES 3/4.6.1.'4 INTERNAL PRESSURE-The limitations on containmcnt internal pres 3ure ensure-that 1) the
. containment structure is prevented from exceeding it design negative pressure-differential with respect to the outside atmosphere of 3.0 psig.
and 2) the containment peak pressure docs not_ exceed the design pressure of 50 psig during LOCA or steam line break conditions.
l The maximum peak pressure expected to be obtained fram a LOCA event is 47.6 psig assus,ing an initial containment pressure of 14.7 psia.
The l.
limit of 1.8-psig for initial positive containment pressure-will limit the total pressure to 49.4 psig which is less than _the design pressure and is consistent with the. accident analyses. The. maximum peak pressure expected to be'obtained from a steam line break event is 49.2 psig assuming an initial containment pressure of 16.5 psia (1.8 psig).
3/4.6.1.5 AIR TEMPERATURE The limitation on containment average air temperature ensures that the containment peak air temperature does not exceed the design 0
' temperature of 276 F during LOCA conditions. The containment temperature limit is consistent with the accident analyses.
3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY The limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility.
Structural integrity is required to ensure that the containment will withstand the maximum pressure of 47.6 psig in the event of a'LOCA. The measurement of containment tendon lift off force, the visual and metallurgical examination of tendons, anchorages and liner and the Type A leakage tests are sufficient to demonstrate this capability.
The-surveillance requirements for demonstrating.the containment's structural integrity are consistent with the intent of.the recommendations of Regulatory Guide 1.35 " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures",
January 1976.
The end anchorage concrete exterior surfaces are checked visually for indications of abnormal material behavior during tendon surveillance.
Inspections of pre-selected concrete crack patterns are performed during the Type A containment leakage rate tests, consistent with the Structural Integrity Test.
Revised by NRC Letter dated May 23. 1990 CM. VERT CiIFFS - UNIT 1 B 3/4 6-2 Amendment No. JSJI, 1
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b, LCONTAINMENT SYSTEMS
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BASES-3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that 1) the
-containment structure is prevented from exceeding it' design negative pressure differential with respect to the outside atmosphere of 3.0 psig K
- and ?)'the containment peak pressure does not exceed the design pressure of 50 psig during LOCA or steam line break conditions.
The maximum peak pressure expected to be obtained from a LOCA event is 47.6' psig assuming an initial containment pressure of 14.7 psia. The l.
limit of 1.8 psig for_ initial positive containment pressure will limit the total pressure to 49.4 psig which is less than the design pressure and-is consistent with the accident analyses.
The maximem peak pressure expected to be obtained from a steam lina break event is 49.2 psig; assuming an initial containment pressure of 16.5 psia (1.8 psig).
3/4.6.1.5 AIR TEMPERATURE The limitation on containment average air temperature ensures that the containment peak air temperature does not exceod the design 0
temperature of 276 F during LOCA conditions. The containment temperature 1
limit is consistent with the accident analyses.
RG l.6 CONTAINMENT STRUCTURAL INTEGRITY The limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards' for the-life of the facility.
Structural integrity is required to ensure that the containment will withstand the maximum pressure of 47.6 psig in the vent of a LOCA. The measurement of containment tendon lift off force, the visual and metallurgical examination of tendons, anchorages and liner and-the Type A leakage tests are sufficient to 1
demonstrato this capability.
The surveillance' requirements for demonstrating the containment's structural integrity'are consistent with the intent of the recommendations of Regulatory Guide 1.35 " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures",
January 1976.
The end anchorage concrete exterior surfaces are checked visually for. indications of abnormal material behavior during tendon surveillance.
Inspections of pre-selected concrete crack patterns are performed during the Type A containment leakage rate tests, consistent with the Structural Integrity Test.
Revised by NRC Letter dated May 23. 1990 CALVERT CLIFFS - UNIT 2 B 3/4 6-2 Amendment No. E,
5'I CONTAINMENT SYSTEMS 4-r BASES 3/4.6.117 CONTAINMENT PURGE SUPPLY AND EXHAUST ISOLATION VALVES
. This limitation ensures that containment purge supply and exhaust valves will be maintained shut during MODES where containment pressurization may occur as the result of LOCA or steam line break conditions. The capability of these valves to close during a containment pressurization event and provide isolation of these _ lines has not been M
established.
3/4.6.2 -DEPRESSURIZATION AND COOLING SYSTEt$
3/4.6.2.1 CONTAINMENT SPRAY SYSTEM s
The OPERABILITY of the containment spray system ensures that containmen_t depressurization and cooling capability will be available in the event of a LOCA.- The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the accident analyses.
a 3/4.6.2.2 CONTAINMENT COOLING SYSTEM i
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The OPERABILITY of the containment cooling system ensures that 1) the containment air temperature will be maintained within limits during normal operation, and 2) adequate heat removal capacity is available during post-LOCA conditions.
s 2
3/4.6.3 I0 DINE REMOVAL SYSTEM The OPERABILITY of the containment iodine filter trains ensures that sufficient iodine removal capability will be available in the event of aL LOCA. Ine reduction in containment iodine inventory reduces the resulting site boundary radiation doses associated with containment leakage.. The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA analyses.
3/4.6.4' CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valv% ensure that the containment atmosphere will be isolated from the cutside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment.
Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.
E s
s Revised by NRC Letter dated Mav 23. 1990 i
CALVERT CLIFFS - UNIT 1 B 3/4 6-3
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CONTAINMENT SYSTEMS' i
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~ BASES 3/4.6.I'.7 CONTAINMENT PURGE SUPPLY AND EXHAUST ISOLATION VALVES' This limitation ensures that containment purge supply and exhaust valves will be.naintained shut during MODES where containment pressurization may occur as the result of LOCA or steam line break conditions. -The capability of these valves to close during a containment pressurization event and provide isolation of these lines has not been established.-
3/4.6.2' DEPRESSURIZATION AND COOLING SYSTEMS a.-
3/4.6.2.I CONTAINMENT SPRAY SYSTEM The OPERABILITY of the containment spray system ensures that
. containment depressurization and cooling capability-will be available in-the event of a LOCA. The pressure reduction and resultant lower containment leak. age rate are consistent with the assumptions used in the accident. analyses.
3/4.6.2.2 CONTAINMENT COOLING SYSTEM The OPERABILITY.of the containment cooling system ensures that'l)
L the containment air. temperature will be maintained within limits during normal operation, and 2) adequate heat removal capacity is available L
'during post-LOCA conditions.
L 3/4.6.3 IODINE REMOVAL SYSTEM
.The OPERABILITY of the containment iodine filter trains ensures. that L
sufficient iodine removal capability will be available in the event of a LOCA.. The reduction in containment iodine inventory reduces the resulting site boundary radiation doses associated-with containment leakage. The operation of this system and resumnt iodine removal-capacity are consistent with the assumptions una in the LOCA analyses.
3/4'.6.4~ CONTAINMENT ISOLATION VALVES l
The OPERABILITY of the containment isolation valves ensure that the containment atmosphere will be isolated from the outside environment in
.the event of a release of radioactive material to the containment atmosphere or pressurization of the containment. Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for c LOCA.
Revised by NRC Letter dated Mw 23.1990 1
CALVERT CLIFFS - UNIT 2 B 3/4 6-3
-aA c
E0WER DISTRIBUTION LIMITS
- . c' -
BASES and Local Power Density. High LCOs and LSSS setpoints remain valid. An AZIMUTHAL POWER TILT > 0.10 is not expected and if it should occur, subsequent operaticn would be restricted to only those' operations required to identify the cause of this unexpected tilt.
thatmustbeusedintheequationFfy.-Fxy (1 + T )
The value of Tq q
andFf-Fr (1 + T ).5 the measured tilt.
.q ThesurveillancerequirementsforverifyingthatFfy,FfandT are q
within their limit; provide assurance that the actual values of Ffy, Ff and T do.not exceed the assumed value. VerifyingFfy,Ffaftereach q
fuel loading prior to exceeding 75% of RATED THERMAL POWER provides additional assurance that the core was properly lorded.
3/4.2.5' DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of
- operation assumed in the transient'and-accident analyses. The limits are consistent with the safety analyses assumptions and have been analytically demonstrated adequate to maintain a minimum DNB Specified Acceptable. Fuel Design Limit. (SAFDL) of 1.15 throughout each analy:ad transient.
In addition to the DNB criterion, there are two other criteria which set the specification in Figure 3.2-4.
The second criterion is to ensure that the' existing core power distribution at full power is less severe thanL the power distribution factored into the small-break LOCA analysis.
TM results in a limitation on the allowed negative AXIAL SHAPE INDEX
. e at full power. The third. criterion is to maintain limitations on
- K linear heat rate at low power levels resulting from Anticipated c,.arational Occurrences (A00s).
Figure 3.2-4 is used to assure the LHR criteria for this condition because the linear heat rate LCO, for both ex-core and in-core monitoring, is set to maintain only the LOCA kw/ft requirements which are limiting at high power levels. At reduced power levels, the kw/ft requirements of certain A00s (e.1,., CEA withdrawal),
tend to become more limiting than that for LOCA.
.The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument readou' is sufficient to ensure that the parameters are 1 stored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.
Revised by HRC Letter dated May 23. 1990 CALVERT CLIFFS - UNIT 1 B 3/4 2-2 Amendment No. 19/fS/55/77/Jpp/
19D9/59
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