ML20045B416
| ML20045B416 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 06/08/1993 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20045B414 | List: |
| References | |
| NUDOCS 9306170325 | |
| Download: ML20045B416 (27) | |
Text
. __ -
ATTACHMENT A-1 i
Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 209 I
l 2
The following is a list of the affected page-1 1
Affected Page:
6-22 l
5 t
I
-h r
l 4
i r
l j
l
-)
I i
9306170325 930608
{
PDR ADOCK 05000334 a
i P
PDR l
l a ~= - - -. -. - -.. ~,,.
i DPR-66 ADMINISTRATIVE CONTROLS SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued)
The radioactive effluent release report to be submitted 60 days i
after January 1
of each year shall also include an assessment of I
j radiation doses to the likely most exposed real individual from l
reactor releases for the previous calendar year to show conformance l
with 40 CFR
- 190, Environmental Radiation Protection Standards for
[
Nuclear Power Operation.
Acceptable methods for calculating the i
dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Revision 1.
The SKYSHINE code (available from Radiation Shielding Information Center, ORNL) is acceptable for calculating the dose contribution from direct radiation due to N-16.
i The radioactive effluent release reports shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each calendar quarter as outlined in 1
In addition, the unrestricted area boundary I
maximum noble gas gamma air and beta air doses shall be evaluated.
The assessment of radiation doses shall be performed in accordance j
vith the ODCM.
The radioactive effluent release reports shall also include,any licensee initiated changes to the ODCM made during the 6 month period.
CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any l
l remaining part of a
reload cycle. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:
1 1.
" WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY",
July 1985 (Westinghouse Proprietary).
Methodology applied for the following Specifications:
1 3.1.3.5, Shutdown Rod Insertion Limits 3.1.3.6, Control Rod Insertion Limits RERJfE 3.2.1, Axial Flux Difference-Constant Axial Offset Control w i7 H 3.2.2, Heat Flux Hot Channel Factor-FQ(Z)
!j 73gggf g 3.2.3, Nuclear Enthalpy Rise Hot Channel Factor-FN delta H
.i I
2.
WCAP-9220-P-A, Rev.
1,
" WESTINGHOUSE ECCS EVALUATION MODEL-1981 f'
SION",
February 1982 (Westinghouse Procrietary).,
Methodology applied for the following spec 1rication:
3.2.2, Heat Flux Hot Channel Factor-FQ(Z) j 3.
" POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING TOPICAL REPORT",
September 1974 (Westinghouse PROCEDURES Proprietary).
Methodology applied for the following Specification:
3.2.1, Axial Flux Difference-Constant Axial Offset Control BEAVER VALLEY - UNIT 1 6-22 Amendment No.
(Propold LDordu
i Attachment to Administrative Controls Insert "A" l
l WCAP-10266-P-A Rev.
2 / WCAP-11524-NP-A Rev.
2, "The 1981 Versic-.O l
the Westinghouse ECCS Evaluation Model Using the BASH Code," K % ?
J.
N.,
et al.,
March 1987; including Addendum 1-A " Power 't3p; Sensitivity Studies" 12/87 and Addendum 2-A
" BASH Methoc o l air,;
Improvements and Reliability Enhancements" 5/88.
i 4
)
J i
i i
1 i
i i
l I
BEAVER VALLEY - UNIT 1 (Proposed Wording) i
1 ATTACHMENT A-2 Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Change No. 76 j
1
)
The following is a list of the affected page:
Affected Page:
6-18
)
l
-i 4
'l 1
1 i
t i
4
't i
l i
i I
i l
I 1
1 j
1
NPF-73 ADMINISTRATIVE CONTROLS SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued)
The radioactive effluent release report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed real individual from reactor releases for the previous calendar year to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation.
Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Revision 1.
The SKYSHINE Code (available from Radiation Shielding Information Center, (ORNL) is acceptable for calculating the dose contribution from direct radiation due to N-16.
The radioactive effluent release reports shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each calendar quarter as outlined in Regulatory Guide 1.21.
In addition, the unrestricted area boundary maximum noble gas gamma air and beta air doses shall be evaluated.
The anessment of radiation doses shall be per-formed in accordance with ODCH.
The radioactive effluent release reports shall also include any licensedt initiated changes to the ODCH made during tne 6 month period.
CORE tdERATING LIMITS REPORT 6.9.1.14 ~ Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:
1.
WCAP-9272-P-A, " WESTINGHOUSE P.ELOAD SAFETY EVALUATION METHODOLOGY", July 1985 (Westinghouse Proprietary). Methodology applied for the following Specifications:
3.1.3.5, Shutdown Rod Insertion Limita 3.1.3.6, Control Rod Insertion Limits 3.2.1, Axial Flux Difference-Constant Axial Offset Control UN TN 3.2.2, Heat Fit:x Hot Channel Factor-F (Z)
INSEPT 6" 9
3.2.3, Nuclear Enthsipy fsise Hot Channel Factor-FN delta H 2.
[WCAP-9220-P-A,Rev.1,"WESTINGHOUSEECCSEVALUATIONM00EL-1981 VERSION",)
Rebruary 1982 (Westinghouse proprietary Methodology applied for the following 5pecification:
3.2.2, Heat F ux Hot Channel Factor-F (Z) 9 3.
WCAP-8385, " POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCECURES -
TOPICAL REPORT", September 1974 (Westinghouse Proprietary).
Methodology applied for the following Specification;_3.2.1, Axial Flux Difference-Constant Axial Offset Control.
4.
T. M. Anderson to K. Kniel (Chief of Core Performance Branch, NRC)
January 31, 1980 --
Attachment:
Operation and Safety Analysis Aspects of an Improved Load Follow Package.
Methodology applied for the following Specification:
3.2.1, Axial Flux Difference-Constant Axial Offset Control.
BEAVE4 VALLEY - UNIT 2 6-18 Amendment No.
gPr pese5 %&
~
Attachment to Administrative Controls i
Insert "B"
WCAP-10266-P-A Rev.
2 / WCAP-11524-NP-A.Rev.'2, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," Kabadi,-
J.
N.,
et al.,
March 1987; including Addendum 1-A " Power Shape i
Sensitivity Studies" 12/87 and Addendum 2-A
" BASH Methodology-
[
Improvements and Reliability Enhancements" 5/88.
'{
i i
i r
i l
1
+
b i
i t
f i
k i,
BEAVER VALLEY-- UNIT 2 (Proposed Wording) i
~
I l
ATTACEMENT B Beaver Valley Power Station, Unit Nos. 1 and 2 Proposed Technical Specification Change No. 209 and 76 REVISION OF SPECIFICATION 6.9.1.14 TITLED
" CORE OPERATING LIMITS REPORT" A.
DESCRIPTION OF AMENDMENT REQUEST The proposed change request would revise Specification 6.9.1.14 titled
" Core Operating Limits Reports" under Administrative Controls.
The specific revision would be to change item number 2 to reference WCAP-10266-P-A Rev.
2 and WCAP-11524-NP-A Rev. 2 instead of WCAP-9220-P-A Rev.
1.
B.
BACKGROUND Generic Letter (GL) 88-16 titled
" Removal of Cycle-Specific Parameter Limits From Technical Specifications" provided guidancs for the preparation of a
license amendment request to modify technical specifications (TS) that have cycle-specific parameter limits.
As part of GL 88-16 implementation, Spec!.fication 6.9.1.14, Core Operating Limits Report (COLR) was added to the reporting requirements of the Administrative Controls Section of the technical specifications.
The COLR provides the values of cycle specific parameter limits that are applicable for the current fuel cycle.
Specification 6.9.1.14 also requires that the values of these limits be established using NRC-approved methodology in topical reports and NRC letters approving a
plant-specific methodology submittal, and consistent with all applicable limits of the safety analysis.
This specification-also requires that all changes in cycle-specific parameter limits be documented in the COLR before each reload cycle or remaining part of a
reload cycle and submitted upon issuance to the NRC, prior to operation with the new parameter limit.
As part of our technical specification change to implement GL 88-16, Request Nos.
167 and 29 submitted on December 14, 1989 for Beaver Valley Power Station (BVPS) Unit No. 1 and 2 respectivesy. We submitted the current list of analytical methods specified in Specification 6.9.1.14 items 1 through 5.
GL 88-16 was implemented on April 26, 1990 through license amendments 154 and 31 for BVPS Unit Nos. 1 and 2 respectively.
The NRC approved methodologies which can be used to determine the Heat Flux Hot Channel Factor F9(Z), referenced in Specification 6.9.1.14, item number 2,
are as follows in chronological order:
WCAP-9220 titled
" Westinghouse ECCS Evaluation Model February 1978 Version,"
WCAP-9220 Revision 1
titled " Westinghouse ECCS Evaluation Model-1981 Version,"
WCAP-9561 BART-Al titled "A
Computer Code For The Best Estimate Analysis Of Reflood Transients,"
and WCAP-10266 Revision 2 titled "The 1981 Version Of The Westinghouse ECCS Evaluation Model Using The BASH Code."
ATTACHMENT B, continued Proposed Technical Specification Change Nos. 209 and 76 Page 2 C.
JUSTIFICATION For BVPS Unit No.
1, the proposed change is editorial in nature.
(
The current item number 2,
under Specification 6.9.1.14, references a WCAP which uses the 1981 version of the Westinghouse ECCS Evaluation Model.
This methodology was incorrectly l
referenced in our Change Request No. 167 submitted December 14, 1989.
BVPS Unit No. 1 first changed from tha 1978 version of the Westinghouse ECCS Evaluation Model to the' BASH Model in December 1987 when the upflow conversion was performed to the reactor vessel.
WCAP-11639 titled "Upflow Conversion Safety Evaluation Report Beaver Valley Unit 1"
documented the use of the BASH l
methodology.
This WCAP was submitted to the NRC on December 7, 1987 in a
letter titled
" Sixth Refueling Outage Plant Modifications Additional Information."
When Technical I
Specification Change Request No.
167 was prepared to add the references to Specification 6.9.1.14 in accordance with GL 88-16, the current methodology (i.e.,
BASH) was not referenced.
- Instead, the predecessor to the BART methodology (i.e.,
the 1981 Version of the Westinghouse ECCS Evaluation Model) was referenced.
Since both methodologies (i.e.,
1981 Version and i
BASH) are approved by the NRC, we are requesting an editorial' change to Specification 6.9.1.14 to reference 'the BASH model which is our current methodology.
For BVPS Unit No.
2, we are requesting a change to item number 2 i
under Specification 6.9.1.14 to allow the utilization of the BASH i
methodology.
The BASH methodology, submitted in WCAP-10266, was approved by the NRC on November 13, 1986.
The current item number 2,
under Specification 6.9.1.14, references a WCAP which uses the "981 Version of the Westinghouse ECCS Evaluation Model.
This methodology was incorrectly referenced in our Change Request No.
29 submitted on December 14, 1989.
BVPS Unit No. 2 has used i
the BART methodology, contained in WCAP-9561-P-A, since before commercial operation.
This fact is docum?nted in Amendment No.
13, dated January 1987, to our Final Safety Analysis Report.
Therefore, Specification 6.9.1.14, item number 2,
should j
reference our current methodology (i.e.,
BART) contained in WCAP-9561-P-A.
l By changing evaluation
- models, BVPS Unit No. 2 will be able to perform Large Break Loss of Ccolant Analysis (LBLOCA) using the i
BASH methodology.
This will result in a lower cumulative peak s
clad temperature (PCT),
currently at 2166*F, and provide margin which will allow increasing the steam generator tube plugging limit.
The changing of evaluation model from the BART to BASH methodology will be handled under the 10 CFR 50.59 process, contingent on approval of this request, in conjunction with the evaluation which will allow increasing the steam generator tube plugging limit.
[
B-2
?
ATTACHMENT B, continued Proposed Technical Specification Change Nos. 209 and 76 Page 3 4
Therefore, we are requesting that the Specification 6.9.1.14 item i
number 2
be revised to reflect the current methodology used at BVPS Unit No. 1 and to allow the use of BASH nethodology at BVPS i
Unit No.
2.
This change is justified based on the change will correct an error in the BVPS Unit No. 1 Technical Specifications' 4
and is editorial in nature.
The change to BVPS Unit No. 2 i
Technical Specifications will allow the -use of a NRC approved methodology and the actual model change will be performed under j
the 10 CFR 50.59 process.
i
~
D.
SAFETY ANALYSIS The proposed revision to Specification 6.9.1.14 for BVPS Unit No.
1 is editorial and doe % not affect the safety of the plant.
i The proposed addition of kne reference to the BASH methodology will reflect the correct methodology that was utilized prior to the addition of the COLR and associated references to Specification 6.9.1.14.
In addition, the BASH methodology has been reviewed and approved by the NRC.
Therefore, the proposed revision to BVPS Unit No. 1 does not affect plant safety.
The proposed revision to Specification 6.9.1.14 for BVPS Unit V
No.
2 will allow the use of the BASH methodology-instead of the BART methodology.
The BASH methodology contained in WCAP-10266 i
has been reviewed and approved by the NRC..
The BASH methodology t
represents the combining of two NRC approved methodologies, BART and NOTRUMP.
As stated in the Safety Evaluation Report (SER) for WCAP-10266, the BASH methodology provides an improved level of realism combined with an acceptable level of conservatism, such that the level of safety and margin of conservatism has been retained.
Therefore, the proposed change is considered safe based on the fact that the BASH methodology will provide the same level of safety and margin of conservatism as the BART and NOTRUMP methodologies which are currently used at BVPS Unit No. 2 to perform the LBLOCA analyses.
The change to reference the BASH methodology for BVPS Unit No. 1 is editorial in nature and does not affect plant safety.
E.
NO SIGNIFICANT HAZARDS EVAI'1ATION The no significant hazard considerations involved with the proposed amendment have been evaluated, focusing on the three standards set forth in 10 CFR 50.92(c) as quoted below:
i I
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 significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not:
S B-3 I
i ATTACHMENT B, continued Proposed Technical Specification Change Nos. 209 and 76 i
Page 4 (1)
Involve a
sinnificant 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 previously. evaluated; or (3)
Involve a significant reduction in a margin of safety.
The following evaluation is provided for the no significant hazards consideration standards.
+
1.
Does
.the change involve a
significant increase in the probability or consequences of an accident previously evaluated?
H For BVPS Unit No.
1, the proposed change to Specification 6.9.1.14 to reference the BASH methodology is editorial in nature since this change does not represent a model change.
Therefore, since this change is editorial, it does not increase the probability or consequences of an accident previously evaluated.
For BVPS Unit No.
2, the proposed change to Specification 6.9.1.14 to reference the BASH methodology will allow the plant to utilize a
NRC approved methodology.
The actual plant implementation will be performed under the 10 CFR 50.59 process.
Since the BASH methodology has been 1
determined to be safe through previcus NRC reviews, the proposed change to the reference in Specification 6.9.1.14 does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2.
Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not change the plant configuration in a
way which introduces a
new potential hazard to the plant.
Since design requirements continue to be met and the integrity of the reactor coolant system pressure is not challenged, no new failure mode has been created.
As a
- result, an accident which is different than any already evaluated in the Updated Final Safety Analysis Report will not be created due to this change.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any j
accident previously evaluated.
l i
B-4 1
~
i ATTACHMENT B, continued Proposed Technical Specification Change Nos. 209 and 76 Page 5 j
3.
Does the change involve a significant reduction in a margin of safety?
The proposed change to reference the BASH methodology in the BVPS Unit No.
1 Technical Specifications is editorial in.
nature.
BVPS-Unit No.
1 currently utilizes the BASH I
methodology to perform the LBLOCA analysis.
The current reference to the 1981 Version of the Westinghouse ECCS i
Evaluation Model methodology was incorrectly added when Specification 6.9.1.14 was revised in accordance with l
Therefore, the proposed revision to reference the l
BASH methodology does not affect the margin of safety at BVPS Unit No.
1.
The proposed change will allow BVPS Unit No. 2 to use the l
BASH methodology.
This change will provide an improved level of realism combined with an acceptable level of conservation for the LBLOCA analysis, such that the level of safety and margin of conservatism has been retained.
l Therefore, the proposed revision to allow the use of the h
BASH methodology will provide the same level of safety and margin of conservatism as the current methodologies used at BVPS Unit No. 2 to perform LBLOCA analyses.
The use of the BASH methodology, therefore, does not involve a significant reduction in the margin of safety.
F.
NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the considerations expressed above, it is concluded that 1
the activities associated with this license amendment request satisfies the no significant hazards consideration standards of 10 CFR
- 50. 92 (c)
- and, accordingly, a
no significant hazards
[
consideration finding is justified.
G.
UFSAR CHANGES Attachment D
provides changes to the UFSAR to accommodate the
~
proposed revision to reference the BASH methodology.
The UFSAR changes are provided for information only and will be incorporated following approval of the proposed Technical Specification changes.
j i
B-5 i
I
P I
ATTACHMENT C-1 i
Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change _No. 209 Applicable Typed Pages 4
I h
9 4
J s
4 J
t
)
)
l E
+
t n
~ ~.
ATTACHMENT TO LICENSE AMENDMENT NO.
6 FACILITY OPERATING LICENSE NO. DPR-66 DOCKET NO. 50-334-Replace the following pages of Appendix A, Technical Specifications, with the enclosed pages as indicated.
The revised pages are' i
identified by amendment number and contain vertical lines indicating i
the areas of change.
Remove Insert i.
6-22 6-22 6-22a 6-22a i
I' r
r 1
3 (Proposed Wording).
1
}
DPR-66 ADMINISTRATIVE CONTROLS t
REMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued)
The radioactive effluent release report to be submitted 60 days after January 1
of each year shall also include an assessment of radiation doses to the likely most exposed real individual from reactor releases for the previous calendar year to show conformance with 40 CFR
- 190, Environmental Radiation Protection Standards for Nuclear Power Operation.
Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in i
Regulatory Guide 1.109, Revision 1.
The SKYSHINE code (available from Radiation Shielding Information Center, ORNL) is acceptable for calculating the dose contribution from direct radiation due to N-16.
The radioactive effluent release reports shall include an assassment I
of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each calendar quarter as outlined in Regulatory Guide 1.21.
In addition, the unrestricted area boundary maximum noble gas gamma air and beta air doses shall be evaluated.
The assessment of radiation doses shall be performed in accordance with the ODCM.
r The radioactive effluent release reports shall also include any licensee initiated changes to the ODCM made during the 6 month period.
CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a
reload cycle.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:
1.
" WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY",
July 1985 (Westinghouse Proprietary).
Methodology applied for the following Specifications:
I 3.1.3.5, Shutdown Rod Insertion Limits 3.1.3.6, Control Rod Insertion Limits i
3.2.1, Axial Flux Difference-Constant Axial Offset Control 3.2.2, Heat Flux Hot Channel Factor-F9(Z) l 3.2.3, Nuclear Enthalpy Rise Hot Channel Factor-FN delta H l
2.
WCAP-10266-P-A Rev. 2 / WCAP-11524-NP-A Rev.
2, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code,"
- Kabadi, J.
N.,
et al., March 1987; including Addendum 1-A " Power i
Shape Sensitivity Studies" 12/87 and Addendum 2-A
" BASH i
Methodology Improvements and Reliability Enhancements" 5/88.
Methodology applied for the following Specification:
3.2.2, Heat Flux Hot Channel Factor-Fg(Z)
BEAVER VALLEY - UNIT 1 6-22 Amendment No.
i (Proposed Wording)
d DPR-66 ADMINISTRATIVE ROLS l
t CORE OPERATING LIMITS REPORT (Continued) 3.
" POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING
)
PROCEDURES TOPICAL REPORT",
September 1974 (Westinghouse Proprietary).
Methodology applied for the following Specification:
3.2.1, Axial Flux Difference-Constant Axial
{
Offset Control
}
4.
T.
M.
Anderson to K. Kniel (Chief of Core Performance Branch, NRC)
January 31, 1980
Attachment:
Operation and Safety Analysis Aspects of an Improved Load Follow Package.
Methodology applied for the following specification:
3.2.1, Axial Flux Difference-Constant Axial Offset Control 5.
NUREG-0800, Standard Review
- Plan, U.
S. Nuclear Regulatory Commission, Section 4.3, Nuclear
- Design, July 1981.
Branch j
Technical Position CPB 4.3-1, Westinghouse. Constant Axial l
Offset Control (CAOC),
Rev.
2, July 1981.
Methodology applied
- l for the following specification:
3.2.1, Axial Flux Difference-Constant Axial Offset Control
.l The core operating limits shall be determined so that all applicable limits (e.g.,
fuel thermal-mechanical limits, core thermal-hydraulic
- limits, ECCS
- limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are i
met.
The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto,- shall be provided on issuance, for each reload cycle, to the NRC Document Control Desk.
SPECIAL REPORTS i
6.9.2 Special reports shall be submitted to the U.
S.
Nuclear Regulatory Commission, Document Control Desk, within the time period specified for each report.
These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:
l l
1 BEAVER VALLEY - UNIT 1 6-22a Amendment No.
(Proposed Wording) i
ATTACHMENT C-2 Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Change No. 76 i
I i
Applicable Typed Pages l
i 1
s ll
)
9 1
I r
1 l
i 1
"F
^'
r
_.,w
_m j
i ATTACHMENT TO LICENSE AMENDMENT NO.
FACJ.LITY OPERATING LICENSE NO. NPF-73 DOCKET NO. 50-412 1
Replace the following pages of Appendix A, Technical Specifications,_
i with the enclosed pages as indicated.
The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
t f
Remove Insert 6-18 6-18 6-18a 6-18a i
i l
I i
i l
i i
I (Proposed Wording)
MPF-73 ADMINISTRATIVE CONTROLS SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued)
The radioactive effluent release report to be submitted 60 days after January 1
of each year shall also include an assessment of radiation doses to the likely most exposed real individual from reactor releases for the previous calendar year to show conformance with 40 CFR
- 190, Environmental Radiation Protection Standards for Nuclear Power Operation.
Acceptable methods for calculating the dose l
contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Revision 1. The SKYSHINE Code (available from l
Radiation Shielding Information
- Center, (ORNL) is acceptable for r
calculating the dose contribution from direct radiation due to N-16.
The radioactive effluent release reports shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each calendar quarter as outlined in Regulatory Guide 1.21.
In addition, the unrestricted area boundary t
maximum noble gas gamma air and beta air doses shall be evaluated.
The assessment of radiation doses shall be performed in accordance with ODCM.
The radioactive effluent release reports shall also include any C
licensee initiated changes to the ODCM made during the 6 month period.
CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documentc5 in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a
reload cycle.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:
1.
" WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY",
July 1985 (Westinghouse Proprietary). Methodology applied for the following Specifications:
3.1.3.5, Shutdown Rod Insertion Limits 3.1.3.6, Control Rod Insertion Limits 3.2.1, Axial Flux Difference-Constant Axial Offset Control 3
3.2.2, Heat Flux Hot Channel Factor-Fg(Z) 3.2.3, Nuclear Enthalpy Rise Hot Channel Factor-FN delta H
[
2.
WCAP-1G266-P-A Rev. 2 / WCAP-11524-NP-A Rev.
2, "The 1981 Version i
of the Westinghouse ECCS Evaluation Model Using the BASH Code,"
- Kabadi, J.
N.,
et al., March 1987; including Addendum 1-A " Power Shape Sensitivity Studies" 12/87 and Addendum 2-A
" BASH Methodology Improvements and Reliability Enhancements" 5/88.
Methodology applied for the following Specification:
3.2.2, Heat Flux Hot Channel Factor-F9(Z)
BEAVER VALLLEY - UNIT 2 6-18 Amendment No.
(Proposed Wording)
l
^
i NPF-73 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)-
3.
" POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES TOPICAL REPORT",
September 1974 (Westinghouse l
Proprietary).
Methodology applied for the follo*ging Specification; 3.2.1, Axial Flux Difference-Constant Axial Offset Control.
4.
T.
M. Anderson to K.
Kniel-(Chief of Core Performance Branch,-
i
Attachment:
Operation and Safety NRC)
January 31, 1980 Analysis Aspects of an Improved Load Follow-Package.
Methodology applied for the following Specification:
3.2.1, Axial Flux Difference-Constant Axial Offset Control.
5.
NUREG-0800, Standard Review
- Plan, U.S.
Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July 1981. Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev.
2, July 1981. Methodology applied for the following Specification:
3.2.1, Axial Flux Difference-Constant Axial Offset Control The core operating limits shall be determined so that all applicable limits (e.g.,
fuel thermal-mechanical limits, core thermal-hydraulic
- limits, ECCS
- limits, nuclear limits such as shutdown margin,-and-l transient and accident analysis limits) of the safety analysis'are met.
The CORE. OPERATING LIMITS
- REPORT, including any mid-cycle revisions or supplements
- thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk.
t SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, -Document Control Desk within the time period specified for each report. These reports shall be submitted covering' the activities identified below pursuant to the requirements of the applicable reference specification:
a.
ECCS Actuation, Specifications 3.5.2 and 3.5.3.
j b.
Inoperable Seismic Monitoring Instrumentation, Specification 3.3.3.3.
c.
Inoperable Meteorological Monitoring Instrumentation, Specification 3.3.3.4.
l 1
l BEAVER VALLLEY - UNIT-2 6-18a Amendment No.
(Proposed Wording)
)
i
i-ATTACHMENT D-1 1
Beaver Valley Power Station, Unit No. 1
'I Proposed Technical Specification Change No. 209 j
Applicable UFSAR' Changes Section 14.3
'I Pages:
14.3-51 14.3-54 i
I
\\
f 3
i 5
l I
1 l
l 5
\\
l
e BVPS-1-UPDATED FSAR Rev. 10 (1/92)
References to Section 14.3 1.
- Lee, H., _
Rupprecht, S.D.,
- Tauche, W.D.,
-Schwarz, W.R.,
" Westinghouse Small Break ECCS Evaluation Model -Using the NOTRUMP Code," WCAP-10054-P-A, August 1985.
2.
- Meyer, P.E.,
"NOTRUMP, A
Nodal Transient Small Break and General Network Code," WCAP-10079-P-A, August 1985.
- 3. J.
M.
- Hellman,
" Fuel Densification Experimental Results and Model for Reactor Application",
WCAP-8219, Westinghouse Electric Corporation (October 1973).
4.
F.
M.
- Bordelon, et al.,
"LOCTA-IV Program:
Loss-of-Coolant Transient Analysis",
WCAP-8305, Westinghouse Electric Corporation (June 1974).
5.
" Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors" 10CFR50.46 and Appendix K of 10CFR50.
Federal Register, Volume 39, Number 3, Westinghouse Electric Corporation (January 4, 1974).
6.
F.
M.
- Bordelon, H.
W.
Massie, and T.
A.
Zordan, " Westinghouse ECCS Evaluation Model Summary" WCAP-8339 Westinghouse Electric Corporation (July 1974).
7.
F.
M.
- Bordelon, et al.,
" SATAN-VI Program:
Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant", WCAP-8302 (Proprietary Version),
WCAP-8306 (Non-Proprietary Vercion),
Westinghouse Electric Corporation (June 1974).
8.
WCAP-14245-P, Rev.
2, "Ths 1901 version-Of-ths Usstinghouse
- ECCS-Evaluat-ion-Hode-1-using-BASH 8, 9.
R.
D.
Kelly, et al.,
" Calculational Model for Core Reflooding after a
Loss-of-Coolant Accident (WREFLOOD Code)", WCAP-8170 (Proprietary Version),
WCAP-8171 (Non-Proprietary Version),
Westinghouse Electric Corporation (June 1974).*
10.
F.
M.
Bordelon and E.
T.
- Murphy,
" Containment Pressure Analysis Code (COCO)",
WCAP-8327 (Proprietary Version),
WCAP-8326 (Non-Proprietary Version),
Westinghouse Electric Corporation (June 1974).*
- 11. Johnson, W. J.
(Westinghouse) letter to Murley, T.
E.
(USNRC),
NS-NRC-89-3463, October 5, 1989.
- 12. Deleted by Revision O.
- 13. F.
M.
- Bordelon, et al.,
" Westinghouse ECCS Evaluation Model-Supplementary Information",
WCAP-8471 (Proprietary Version),
WCAP-8472 (Non-Proprietary ' Version),
Westinghouse Electric Corporation (January 1975).
FWCAP-10266-P-A Rev. 2/WCAP-11524-NP-A Rev 2, The 1981 Version of the Westinghouse ECCS Emluation Model Using the BASH Code', Kabadi, J. N., et al., March 1987: including Addendwn 1-A ' Pour Shape Sensitivity Studies' 12/87and Addendum 2-A ' BASH Methdology imprm'ements and Reliabi Enhancements' 5/88 j
4&S['
Q}
BVPS-1-UPDATED FSAR Rev. 10 (1/92)
References to Section 14.3 (Cont'd) l
- 49. Deleted by Revision 2 i
- 50. Deleted by Revision O.
- 51. Deleted by Revision O.
1
- 52. Deleted by Revision O.
- 53. Letter from C.
N. Dunn (Duquesne Light company) to R. W. Reid (Nuclear Regulatory Commission),
Subject:
proposed permanent modifications to correct NPSH modifications.
(November 17, 1977).
- 54. Letter from C.
N.
Dunn (Duquesne Light Company) to A.
Schwencer (Nuclear Regulatory Commission),
Subject:
completed request for additional information (August 3, 1978).
- 55. Deleted by Revision O.
- 56. Deleted by Revision O.
57.
F.
Kreith, Princioles of Heat Transfer, International Textbook Company (1966).
- 58. Deleted by Revision 7.
- 59. " Westinghouse ECCS Evaluation Model 1981 Version",
I WCAP-9220-P-A, Revision 1, February 1982.
Qu c PP- %Q1 - N P-A \\
- 60. Deleted by Revision 7.
Cl
- 61. Deleted by Revision 7.
- 62. Deleted by Revision 7.
- 63. Deleted by Revision 7.
- 64. Deleted by Revision 7.
f
- 65. Deleted by Revision 7.
- 66. Safety Evaluation in the Matter of Virginia Electric Power
- Company, Surry Power Stations Units 1 and 2, Docket Numbers 50-280 and 50-281, pp.
57-58, Atomic Energy Commission (February 23, 1972).
- 67. NUREG-0772,
" Technical Bases for Estimating Fission Product Behavior During LWR Accidents", Appendix E, June 1981.
- 68. DLC Calculation ERS-SFL-83-016 1
- 69. DLC Calculation ERS-SFL-83-017 14.3-54 (l'rspsk
ATTACHMENT D-2 i
Beaver Valley Power Station, Unit No. 2 l
Proposed Technical Specification Change No. 76 l
r Applicable UFSAR Changes l
Section 15.6 Pages:
15.6-14
^
15.6-15 15.6-22 15.6-23 j
t i
I I
.i f
i 6
)
j
i SVPS-2 LTSAR Description of Small Break Loss-of-Coolant Accident Transient As contrasted with the large break, the blowdown phase of the small break occurs over a longer time period. Thus, for the small break LOCA there are only three characteristic stages, that is, a gradual blowdown in which the decrease in water level is checked, core recovery, and long-term recirculation.
15.6.5.3 Core and System Performance 15.6.5.3.1 Mathematical Model The requirements of an acceptable ECCS evaluation model are presented in Appendix K of 10 CFR 50- (Federal Register 1974).
1.
Large Break LOCA Evaluation Model The analysis of a large break LOCA transient is divided into three phases: 1) blowdown, 2) refill, and 3) reflood. There are three distinct transients analyzed in each phase, including the thermal-hydraulic transient in the RCS, the pressure and temperature transient within the containment, and the fuel and clad temperature transient of the hottest fuel rod in the core.
Based on these considerations, a system of interrelated computer codes has been developed for the analysis of the LOCA.
A description of the various aspects of the LOCA analysis methodology is given by Bordelon, Massie, and Zordan (1974).
This document describes the major phenomena modeled, the interfaces among the computer codes, and the features of the codes which ensure compliance with the acceptance criteria. x The SATAN-VI, WREFLOOD, COCO, BAP5,*~and LOCTA-IV codes, OM which are used in the LOCA analysis, are described in detail by Bordelon et al (1974a): Kelly et al (1974) Bordelen and Murphy (1974); houngr-MrY., c t-el-WCAP-9561'-PA1984-) ;
- Chiou, J.S.,
et al, WCAP-10062 (1982); and Bordelon et al (1974b). Code modifications are specified by Bordelen et al' (1975), WCAP-8622 and WCAP-8623 (Westinghouse 1975a). These codes assess the core heat transfer geometry and determine if the core remains amenable to cooling throughout and subsequent t-the blowdown, refill, and reflood phases of the LOCA. The SATAN-VI computer code analyzes the thermal-hydraulic transient in the RCS during blowdown and the WREFLOOD computer code calculates this transient during the refill and reflood phases of the accident.
The COCO computer code calculates the containment pressure transient during all three phases of the LOCA. analysis. Similarly, the LOCTA-IV computer code computes the thermal transient of the hottest fuel rod during the three phases.
fhak: 3, Y. ; O b ' I e
15.6-14 (krpse m.
m.
m-m
BVPS-2 UFSAR SATAN-VI calculates the RCS pressure, enthalpy, density, and the mass and energy flow rates in the RCS, as well as steam generator energy transfer between the primary and secondary systems as a function of time during the blowdown phase of the LOCA.
SATAN-VI also calculates the accumulator water
-l mass and internal pressure and the pipe break mass and i
energy flow rates 'that are assumed to be vented to the containment during blowdown. At the end of the blowdown
- phase, these data are transferred to the WREFLOOD code.
Also, at the end-of-blowdown, the mass and energy release rates during blowdown are transferred to the COCO code for use in the determination of the containment pressure response during this first phase of the LOCA. Additional SATAN-VI output data from the end-of-blowdown, including the core inlet flow rate and enthalpy, the core pressure, and the core power decay transient, are input to the LOCTA-IV code.
1 With input from the SATAN-VI code. WREFLOOD uses a system thermal-hydraulic model to determine the core flooding rate (that is, the rate at which coolant enters the bottom of the core), the coolant pressure and temperature, and the quench front height during the reflood phase of the LOCA. WREFLOOD also calculates the mass and energy flow addition to the containment through the break. Since the mass flow rate to the containment depends upon the core flooding' rate and the local core pressure, which is a function of the containment back-pressure, the WREFLOOD and COCO codes are interactively linked.
The WREFLOOD output is then used as input and g g g _ / boundary conditions forNAM.
The mechanistic core heat J_
transfer model in*'sART then calculates the fluid and heat sfer conditions in the core during reflood.
WREFLOOD i
and output is then transferred by tape to the LOCTA-IV code. LOCTA-IV is used throughout the analysis of the LOCA transient to calculate the fuel clad temperature and metal-water reaction of the. hottest rod in the core.
The large break analysis was performed with the b oomber 1981 ver+1-on--of-the-Evaluet4erHWh+r-with-BAM
("ct;ng, et
.1, 193')
uhich incledes.medificetiss Allu oicd by Eicheidinger---(l o RI )
=nd D a a /toA?).
The large break analysis was performed w'
he December 198 of the ECCS E model as described above and incor
+ed changes necessary to model the pla nfiqUration with one loop
_rvice.
~
Thes nges are delineated in WCAP-8904 (1979).
WCAP-10266-P-A Rev. 2/WCAP-11524-NP-A Rev 2, 'The 1981 Version of the Westinghouse)\\
\\
ECCS Emluation Model Using the BASH Code', Kabadi, J. N., et al., March 1987; including
\\
Addendum 1-A ' Power Shape Sensitivity Studies' 12/87 and Addendum 2 A ' BASH Methdology, Improvements and Reliability Enhancements' 5/88.
' 15.6-15 wp
BVPS-2 LTSAR these sources.
Parameters required to calculate the control room doses are provided in Tables 15.6-11 through 15.6-14.
The total doses at the exclusion area boundary and the LPZ, presented in Table 15.0-12 are within the guidelines of 10 CFR 100.
The dose to the BVPS-2 control room operators due to a LOCA at the BVPS-2 plant, as presented in Table 15.0-13, is below the limit set in General Design Criterion 19 of 5 Rem whole body, or its equivalent to any part of the body.
15.6.6 Boiling Water Reactor Transients Not applicable to BVPS-2.
15.6.7 References for Section 15.6
- Bordelen, F.
M.
et al 1974a.
LOCTA-IV Program: Loss-of-Coolant Transient Analysis.
WCAP-8301 (Proprietary) and WCAP-8305 (Non proprietary).
- Bordelon, F.
M. et al 1974b. SATAN-VI Program Comprehensive Space Time Dependent Analysis of Loss-of-Coolant. WCAP-8302 (Proprietary) and WCAP-8306.
P Bordelen, F. M.; Massie, H. W.; and Zordan, T. A. 1974. Westinghouse ECCS Evaluation Mode - Summary. WCAP-8339.
- Bordelon, F. M.
and Murphy, E. T.
19*. 4.
Centainment Pressure Analysis Code (COCO) and WCAP-8327 (Proprietary) and WCAP-8326 (Non-proprietary).
Bordelon, F.M. et al 1975.
Burnett, F. W. F. et al 1972. LOFTRAN Code Description, WCAP-7907.
- Chieu, 3.
S. et al 1482.
odel: for T"" R ficed Calculstiens ning the-B ART-Code, -WCAkl0062, March--MM.
I Eieheidinger, C.
1981.
Yes tin;heur e ECCS E"elustica Mode, 1981 Veenen.
"C.'.P-9220-P * (Pr4prietary vertien), YCAP-9221-P n (Non-propric t ry er:icr.), n: vision-1.
Federal Register 1974.
Acceptance Criteria for Emergency Core Cooling System for Light Water Cooled Nuclear Power Reactors.
10 CFR 50.46 and Appendix K of 10 CFR 50.
Volume 39, Number 3.
- Johnson, W. J.;
- Massie, H. W.;
and
- Thompson, C. M.
1975.
Westinghouse ECCS-Four Loop Plant (17 x 17) Sensitivity Studies.
WCAP-8586-P-A (Proprietary) and WCAP-8566-A (Non-proprietary).
hrsge3e m a 5.6-22 1
WCAP-10266-P-A Rev. 2/WCAP-11524-NP-A Rev 2, The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code', Kabadi, J. N., et al., March 1987; including Addendwn 1-A ' Power. Shape Sensitivity Studies' 12/87 and Addendum 2-A ' BASH Methdology, Imprmements and Reliability Enhancements' S/88.
o BVPS-2 UFSAR
- Johnsen, W.
J.. and Thompson, C. M.
1977 Westinghouse Emergency Core Cooling System Evaluation Model - Modified October 1975 version.
l WCAP-9168 (Proprietary) and WCAP-9169 (Non-proprietary).
- Kelly, R. D.
et al 1974.
Calculational Model for Core Reflooding after a Loss-of-Coolant Accident (WREFLOOD Code).
WCAP-8170 (Proprietary) and WCAP-8171 (Non-proprietary).
- Kemper, R.
M.
1979.
Westinghouse Emergency Core Cooling System Evaluation Model for Analyzing (N-1) Loop Operation of Plants With Loop Isolation Valves. WCAP-8904-A.
- Lee, H.,
Rupprecht, S.
D.,
- Tauche, W.
D.,
- Schwarz, W.
R.,
" Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A. August 1985.
- Meyer, P.
E.,
"NOTRUMP, A Nodal Transient Small Break and General Network Code," WCAP-10079-P-A, August 1985.
Forsching, T. A.;
Mu rphy,
J. H. ; Redfield, J. A. ; and Davis, V. C.
1969.
FLASH-4, A Fully Implicit FORTRAN-IV Program for the Digital Simulation of Transients in a Reactor Plant.
WARD-TM-84, Bettis Atomic Power Laboratory.
- Rahe, E.
P.
(Westinghouse) 1982 (letter dated November 8, 1982) to James R. Miller (USNRC), letter number NS-EPRS-2679.
Salvatori, R.
1974. Westinghouse ECCS - Plant Sensitivity Studies.
WCAP-8340 (Proprietary) and WCAP-8356 (Non-proprietary).
U.S.
Nuclear Regulatory Commission 1975. Reactor Safety Study - An.
I Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants.
WASH-1400 NUREG-75/014.
Westinghouse Electric Corporation (Westinghouse) 1974. Westinghouse ECCS Evaluation Model Sensitivity Studies.
WCAP-8341 (Proprietary) and WCAP-8342 (Non-proprietary).
Westinghouse 1975a.
Westinghouse Mass and Energy Release Data for Containment Design WCAP-8321A Rev. 2.
Westinghouse 1975b.
Westinghouse ECCS Evaluation Model - October 1975 Version.
WCAP-8622 (Proprietary) and WCAP-8623 (Non-proprietary).
Westinghouse 1978.
Personal communication between C. Eicheldinger, Westinghouse and J.F. Stoltz, USNRC, letter dated February 10,
- 1978, letter number NS-CE-1672.
Young--M.
7.
et :1 1984.
E7.RT--A1.
?. Gomputer-Gode-f:-r the 2:st Freimate-Analysis-of--Ref4eed--Transients,---WGA."1551-P-Ar-Mard 1504.
15.6-23
([ropo3cb i