ML20236Y357
| ML20236Y357 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/08/1987 |
| From: | Shelton D TOLEDO EDISON CO. |
| To: | |
| Shared Package | |
| ML20236Y344 | List: |
| References | |
| NUDOCS 8712110263 | |
| Download: ML20236Y357 (13) | |
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J 6 Docket No. 50-346 License No. NPF-3 Serial No. 1436 !
Enclosure l Page 1 .,
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l i l APPLICATION FOR AMEND!fENT !
l TO ,
l FACILITY OPERATING LICENSE NO. NPF-3 !
FOR l i
DAVIS-BESSE NUCLEAR POWER STATION 1
UNIT NO. 1 I l
J Ia l
Attached is a requested change to the Davis-Eesse Nuclect Power Station, 'l Unit No. 1 Facility Operating License No. NPF-3. Also included are the l Safety Evaluation and Significant Hazards Consideration. j The proposed change (cubmitted under cove: letter Serial No. 1436) concern:
Section 3/4.6.1.2, Containment Systems, Containment Leakage, .
Specification 4.6.1.2c.3. I x
By /I D. C. Shelton N]
Vice President, Nuclear Sworn and subscribed before me this 8th day of December 1987 /
/ J A _ f_ { r , Lk.
Notarf Public, expires State of Ohip/
,[
My commiasion 3/ \ /'7 / ,
B712110263 871200 PDR ADOCK 05000346 P PDR
Docket No. 50-346 n License No. NPF-3 Serial'No.'1436 Enclosure Page 2 The ic11oving information is provided to support issuance of the requested change to the Davis-Besse Nuclear Power Station, Unit No. 1
, Operating License No. NPF-3, Appendix A, Technical Specification 4.6.1.2c.3.
L l A. Time required to implement: This change will be implemented by the licensee within 30 days following NRC issuance of the License Amendment.
B. Reason for change (FCR No. 85-0030 Revision A): Revise the Technical Specifications to make the Type A supplemental verification test consistent with:10CFR50 Appendix J and ANSI N45.4-1972.
C. Safety Evaluation: See attached' Safety Evaluation (Attachment No. 1).
D. Summary Significant Hazarde; Consideration: See attached Summary Significant Hazards Consideration (Attachment No. 2).
E. Significant Hazards Consideration: See attached Significant Hazards Consideration (Attachment No. 3).
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Docket No.'50-346
. , License No. NPF-3 Serial'No. 1436 Attachment 1 Page 1 SAFETY EVALUATION {
l INTRODUCTION l
The purpose of this safety evaluation is to review a proposed change to the Davis-Besse Nuclear Power Station (DENPS), Unit No.1 Operating License, Appendix A, Technical Specifications, Section 4.6.1.2c.3, to ensure that no unreviewed safety question exists. This safety evaluation is being performed to meet the requirements of 10CFR50.59.
Per the Davis-Besse Updated Safety Analysis Report (USAR), Section ,
6.2.1.4.1, Tests and Their Purposes, 10CFR50. Appendix J Type A tests are ]
performed to verify the overall integrity of the containment under the i pressure conditions that might occur following a design-basis accident. )
In performing Type A tests, the containment is pressurized to the peak accident pressure, and the loss per unit time is measured by' converting pressure loss to leskage. Measurements during the test include pressure, temperature, and vapor content of the air. Following performance of the Type A test, a verification test is performed. This verification test consists of imposing a known Jeak rate on the containment after the Type A test data has been recorded for test data accuracy.
The current requirement to perform the containment leakage test and .
verification of the accuracy of each Type A test is identified in the i Davis-Besse Technical Specifications, Section 3/4.6.1, Primary Contain- 1 ment, Specification 4.6.1.2, Containment Systems, Containment Leakage, j Surveillance Requirements. This requirement states that the accuracy of j each Type A test shall be verified by a supplemental test and shall be ;
determined in conformance with the criteria specified in Appendix J of j 10CFR50 using the rathods and provisions of ANSI N45.4-1972. j I
Specifically, Technical Specification 4.6.1.2c.3 requires the quantity of I gas injected into the containment or bled from the containment during the supplemental test to be equivalent to at least 25 percent of the total measured leakage rate at P , 38 psig. (P is defined under the " Effects on Safety" section of this asafety evaluation).
This Technical Specification is contrary to the requirements of ANSI N45.4-1972 and is not consistent with the NRC position that the rate at which gas is injected or bled from containment be between 0.75 and 1.25 L. (L is defined under the " Effects on Safety" section of this safety 1 e0aluation). As presently written, if the initial Type A teat leakage 1 was zero, the supplemental test could not be performed, since 25 percent l of zero is zero.
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- Docket No. 50-346
. License No. NPF-3 Serial No. 1436 Attachment 1 Page'2 Accordingly, Technical Specification 4.6.1.2c.3 should be modified as follows:
" Requires that the rate at which gas is injected into the containment-or bled from the containment during the supplemental test is between 0.75 L,and 1.25 L,."
The proposed Technical Specification will make the wording consistent with 10CFR50 Appendix J, ANSI /ANS 56.8-1981 and ANSI N45.4-1972.
SYSTEMS AFFECTED l
Containment vessel and penetrations systems DOCUMENTS AFFECTED Davis-Besse Nuclear Power Station, Unit'No. 1 Operating License, Appendix A,-Technical Cpeciffcations, Section 4.6.1.2c.3 Davis-Besse Nuclear Power Station, Unit No. 1 Updated Safety Analysis Report (USAR), July, 1987, Section 6.2.1.4.3 ;
Operations Procedure Manual Volume OP21, DB-OP-3009 (ST5061.01),
Containment Integrated Leak Rate Test REFERENCES
- 1. Davis-Besse Nuclear Power Station, Unit No. 1 Operating License, Appendix A, Technical Specificaricns, Section 4.6.1.2c.3
- 2. Davis-Besse Nuclear Power Station, Unit No. 1 Updated Safety Analysis Report, July, 1987, Sections 6.2.1.4.1 and 6.2.1.4.3
- 3. Title 10 Code of Federal Regulations Part 50, Appendix J, Primary l Reactor Containment Leakage Testing for Water Cooled Power Reactors, Paragraphs III A.3(a) and (b) 4.- American National Standards Institute (ANSI) N45.4-1972, Leakage Rate Testing of Containment Structures for Nuclear Reactors, Appendix C.
- 5. American National Standards Institute /American Nuclear Society (ANSI /ANS) 56.8-1981, Containment System Leakage Testing Requirements, Section 3.2.6(b)(1) and Appendix C, Supplemental Method 1.
- 6. Title 10 Code of Federal Regulations Part 100, Reacter Site Criteria.
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. Docket No. 50-346
, License No. NPF-3 Serial No. 1436 Attachment 1 Page 3 FUNCTIONS OF AFFECTED SYSTEMS 1he Containment Vessel and Penetration System is designed to provide protection for the public from the consequences of any break in the reactor coolant piping up to and including a double-ended break of the largest reactor coolant pipe assuming unobstructed discharge from both ends. Pressure and temperature behavior subsequent to the accident is determined by the combined influence of the energy. sources, heat sinks and engineered safety features.
The containment. system also provides protection for the public from the radiological consequences of a maximum hypothetical accident discussed in USAR Chapter 15. The containment design, along with the engineered safety features provided, ensure that the exposure of the public resulting
.from a hypothetical accident is below the guidelines established by 10C FR100.
The Containment Vessel was tested at the conclusion of construction and after all penetrations had been installed to verify that the design leakage rate associated with an internal pressure of 38 psig does not exceed 0.5 percent of the containment contained weight of air and vapor in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The analysis in USAR Chapter 15 obows that this is more than adequate to meet the guidelines of 10CFR100.
EFFECTS ON SAFETY The Type A test is intended to measure the primary reactor containment overall integrated leakage rate (1) af ter the containment has been com-pleted and is ready for operation, and (2) at periodic intervals there-af ter.. - The accuracy of Type A tests is verified by a supplemental test as described in 10CFR50, Appendix J, Paragraph III A.3(b):
"The accuracy of any Type A test shall be verified by a supplemental test. An acceptable method is described in Appendix C of ANSI N45.4-1972. The supplemental test method selected shall be conducted for sufficient duration to establish accurately the change in leakage rate between the Type.A test and supplemental test. Results from this supplemental test are acceptable provided the difference between the supplemental test data and the Type A test data is within 0.25 L .-(or 0.25 L ), If results are not within 0.25 L r$ason shall be determined, corrective action take$,(or and 0.25 L ), the a successful supplemental test performed".
L is defined as the maximum allowable leakage rate at pressure P as j s*ecified for preoperational testo in the Technical Specifications or
) associated beses, and as specified for periodic tests in the operating license.
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' Docket No. 50-346 Liceost No. NPF-3 Serial No. 1436 l Actachment 1 l Page 4 i
P is defined as the calculated peak accident containment internal pSessure related to tlie design basis accident and specified either in the Technical Specifications or associated bases (38 psig).
L is defined as the maximum allowable leakage rate at pressure P dErivedfromthepreoperationaltestdata, t P is defined as the containment vessel rtdeced test pressure selected to measure the inregrated leakage rate during periodic Type A tests.
ANSI N45.4-1972, Appendix C, outlines a method for verification of leakage rate tests. The method irvolves the accurate measurement of the leakage rate through a calibrated leak intentionally superimpeaed on the existing leaks in a containment structure.
An arrangement for superimposing a controlled and measurable leak on the I containment vessel employs tha orifice leak of a microadjustable instru-ment flow valve installed at a conyenient penetration of the containment vessel. The flow through the valve is measured by means of a suitable flow meter. The leak orifice is selected to provide a flow under the test pressure conditions approximately equivalent to the leakage rate apecified for the containment vessel.
ANSI /ANS 56.8-1981 provides further guidance on performance of the supple-mental test. It states that the leak orifice is selected to provide a flow under the test-pressure condition equivalent to 75 percent to 125 percent of the leakage rate specified for the containment system (L,).
During performance of the Type A leakage test, if the leakage rate is determined to te zero, Technical Specification 4.6.1.2c.3 could not be performed since 25 percent cf zero is zero. Therefero, since no gas wcald be injected into the containment cr bled from the containment during the supplemental test, the accuracy of the Type A test could not be
. verified. However, if the guidance of ANSI /ANS 56.8-1981 is followed regarding the 75 to 125 percent of the total allowable leakage rate for the containment being bled from the containment during the supplemental test, the accuracy of the Type A test can be verified during zero leakage conditions.
Revising Technical Specification 4.,6.1.2c.3 mckas the Technical Specific-ceions consistent with the requirements of ANSI /ANS 56.8-1981, ANSI N45.4-1972 and current NRC posftion. This proposed change does not affect plant safety and will enhance the acceptability of the Type A test.
UNREVIEWED SAFETY Q'JESTTON FVALUATION The implementation of this proposed change would not increase the probability of occurrence of an accident previous 1f evaluated in the
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Docket No. 50-346
. License No. NPF-3 Serial No. 1436
, Attachment 1 Page 5 USAR because the proposed Technical Specification change only modifies the supplemental test used to verify the accuracy of the Type A test and does not modify or create any accident initiating condition (10CFR50.59 (a) (i)) .
The implementation of this proposed change would not increase the conse-quences of an accident previously evaluated in the USAR because the proposed Technical Specification change only modifies the supplemental test used to verify the accuracy of the Type A test and does not modify or create any accident initiating condition (10CFR50.59 (a)(2)(1)).
The implementation of this proposed change would not increase the proba-bility of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR because the proposed Technical Specifi-cation change does not involve any change to equipment or Type A test (10CFR50.59 (a) (2) (1)) .
The implementation of this proposed change would not increase the conse-quences of a malfunction of equipment important to safety previously evaluar.ed in the USAR bccause the proposed Technical Specification change does not involve any change to equipment (10CFR50. 59 (a) (2) (1)) .
The implementation of this proposed change would not create the pessi-bility for en accident of a different type than any evaluated previously in the USAR because the proposed Technical Specification change only modifies the supplemental test used to verify the accuracy of the Type A test and does not modify or create any accident initiating condition (10CFR50.59 (a) (2) (ii)) .
The implementation of this proposed change would not create the possi-bility for a malfunction of a different type than any evaluated previously in the USAR because the proposed Technical Specification change only modifies the supplemental test used to verify the accuracy of the Type A test and does not modify or create any accident initiating condition ;
(10CFR50.59 (a) (2) (11)) . l The implementation of this proposed change would not reduce the margin of safety as defined in the basis for any Technical Specification because all the assumptions in the USAR analyses are unchanged (10CFR50.59(a)(2)(iii)).
CONCLUSION Based on the above, it is concluded that the proposed Technical Specific-ation change does not create any unreviewed safety question.
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I Dockat No. 50-346 i License No. NPF-3 Serial No. 1436 Attachment 2 Page 1-1 I
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SUMMARY
SIGNIFICANT HAZARDS CONSIDERATION l
l Description of Amendment Request: This amendment request proposes )
modifying the acceptance criteria for the supplemental verification test !
to the Type A Containment Integrated Leak Rate Test (CILRT) from depender.ce l on a measured value to dependence on a constant value. This change is in !
accordance with current NRC position oli the acceptance criteria. This j requirement is contained in Technical Specification 4.6.1.2c.3. !
Basis for Proposed No Significant Hazards Consideration Determination:
The purpose of this change is to be consistent with 10CFR50 Appendix J, ANSI N45.4-1972, and ANSI /ANS 56.8-1981 by modifying Technical Specification 4.6.1.2c.3 to-read, " Requires that the rate at which gas is )
injected into the containment or bled from the containment during the j supplemental test is between 0.75 L and 1.25 3 1,." l The Commission has provided guidance concerning the application of standards for determining whether license amendments involve Significant Hazards Considerations by providing certain examples, published in the Federal Register on March 6, 1986. One of the examples (ii) of an action involving no significant hazards considerations is a' change that constitutes an additional limitation, restriction, or control, e.g., a more stringent surveillance requirement.
While this change does not provide en additional limitation, restriction or control, it does serve to clarify the test requirement and increase its applicability. This change does not decrease the strfugency of the requirement. This change is identical to the corresponding Technical Specification issued for the Palo Verde Nuclear Generating Station - Unit No. 1 (ref erence NUREG-1133, dated May,1985) .
Therefore, it is concluded that the amendment application involves no significant hazards consideration.
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Docket No. 50-346 l License No. NPF-3 Serial No. 1436 L
Attachment 3 Page 1 SIGNIFICANT HAZARDS CONSIDERATION INTRODUCTION The purpose of this Significant Hazards Consideration is to review a propssed change to the Davis-Besse Nuclear Power Station (DBNPS), Unit No.1 Operating License, Appendix A Technical. Specifications, Section 4.6.1.2c.3, to ensure that no significant hazards consideration exists.
If granted, this change will resolve NRC Inspection Report Open Item 50-346/84029-02. This significant hazards consideration is being performed to meet the requirements of 10CFR50.92(c).
Per the Davis-Besse Updated Safety Analysis Report (USAR), Section 6.2.1.4.1, Tests and Their Purposes, 10CFR50 Appendix J Type A tests are performed to' verify the overall integrity of the containment under the pressure conditions that might occur following a design-basis accident.
In performing Type A tests, the containment is pressurized to the peak accident pressure, and the loss per unit time is measured by converting pressure loss.to leakage. Measurements during the test include pressure, temperature, and vapor content of the air. Following performance of the Type A test, a verification test is performed. This supplemental verification test consists of imposing a known leak rate on the containment after the Type A test data has been recorded for test data accuracy.
The current requirement to perform the containment leakage rate test and verification of the accuracy of each Type A test is identified in the Dovis-Besse Technical Specifications, Section 3/4.6.1, Primary Contain-ment, Specification 4.6.1.2, Containment Systems, Containment Leakage, Surveillance Requirements. This requirement states that the accuracy of each Type A test shall be verified by a supplemental test and shall be determined in conformance with the criteria specified in Appendix J of 10CFR50 using the methods and provisions of ANSI N45.4-1972.
Specifically, Technical Specification 4.6.1.2c.3 requires the quantity of gas injected into the containment or bled from the containment during the i supplemental test to be equivalent to at least 25 percent of the total measured leakage rate at P , 38 psig. (P is defined under the " Effects on Safety" section of this"significant haEards consideration).
This Technical Specification is contrary to the requirements of ANSI N45e4-1972 and is not consistent with the NRC position that the rate at which gas is injected or bled from containment be between 0.75 and 1.25 ;
L. (L is defined under the " Effects on Safety" section of this sfgnif10 ant hazards cor.siderati6n) . As presently written, if the initial Type A test leakage was zero, a valid supplemental test could not be performed under the requirements of 10CER50, Appendix J and ANSI N45.4-1972.
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f h Docket No. 50-346 License No. NPF-3 Serial No. 1436 Attachment 3
.Page 2 1
Accordingly, Technical Specification 4.6.1.2c.3 should be modified as follows:
" Requires that the rate at which gas is injected into the containment or bled from'the containment during the supplemental test is between 1 0.75 L,and 1.25 L,."
The proposed Technical Specification will make the wording more consistent with 10CFR50 Appendix J, ANSI /ANS 56.8-1981 and ANSI N45.4-1972, and identical to that issued for the Palo Verde Nuclear Generating Station -
Unit No. 1 (reference NUREG-1133, dated May_1985).
SYSTEMS AFFECTED Containment vessel and penetrations systems DOCUMENTS AFFECTED Davis-Besse Nuclear Power Station, Unit No. 1 Operating License, Appendix A, Technical Specifications, Section 4.6.1.2c.3 Davis-Besse Nuclear Power Station, Unit No. 1 Updated Safety Analysis Report (USAR), July, 1987, Section 6.2.1.4.3 Operations Procedure Manual Volume OP21, DB-0P-3009 (ST5061.01),
Containment Integrated Leak Rate Test REFERENCES
- 1. Davis-Besse Nuclear Power Station, Unit No. 1 Operating License, Appendix A, Technical Specifications, Section 4.6.1.2c.3 l
- 2. Davis-Besse Nuclear Power Station, Unit No. 1 Updated Safety Analysis Report, July, 1987, Sections 6.2.1.4.1 and 6.2.1.4.3 j
- 3. Title 10 Code of Federal Regulations Part 50, Appendix J Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors, j Paragraphs III A.3(a) and (b)
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- 4. American National Standards Institute (ANSI) N45.4-1972, Leakage Rate Testing of Containment Structures for Nuclear Reactors,
-Appendix C.
'5. American National Standards Institute /American Nuclear Society (ANSI /ANS) 56.8-1981, Containment System Leakage Testing Requirements, Section 3.2.6(b)(1) and Appendix C, Supplemental Method 1.
- 6. Title 10 Code of Federal Regulations Part 100, Reactor Site Criteria.
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- 7. Palo Verde Nuclear Generating Station, Unit No. 1, Technical i Specification 4.6.1.2c.3 issued as part of NUREG-1133, dated May, 1985).
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Docket No.-50-346 License No. NPF-3 i Serial No. 1436 l Attachment 3 Page 3 I- ' FUNCTIONS OF AFFECTED SYSTEMS The Containment Vessel and Penetration System is designed to provide !
l protection for the public from the consequences of any break in the reactor coolant piping up to and including a double-ended break of the largest reactor coolant pipe assuming unobstructed diecharge from both ends.' Pressure and temperature behavior subsequent to the accident is determined by the combined influence of the energy sources, heat sinks and engineered safety features. j The' containment system also provides protection for the public from the l radiological consequences of a maximum hypothetical accident discussed !
in USAR Chapter 15. The containment design, along with the engineered !
safety features provided, ensure that the exposure of the public resulting .!
from a hypothetical accident is below the guidelines established by 10CFR100.
i The Containment Vessel was tested at the conclusion of construction and after all penetrations had been installed to verify that the design leakage rate associated with an internal pressure of 38 psig does not exceed 0.5 percent of the containment contained weight of air and vapor in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The analysis in USAR Chapter 15 shows that this is more than adequate to meet the guidelines of 10CFR100, 9 l
EFFECTS ON SAFETY t The Type A test is intended to measure the primary reactor containment overall integrated leakage rate (1) after the containment has been com-pleted and is ready for operation, and (2) at periodic intervals there- ;
after. The accuracy of Type A tests is verified by a supplemental test as described in 10CFR50, Appendix J, Paragraph III A.3(b):
"The accuracy of any Type A test shall be verified by a supplemental test. An acceptable method is described in Appendix C of ANSI N45.4-1972. The supplemental test method selected shall be conducted '
for sufficient duration to establish accurately the change in leakage rate between the Type A test and supplemental test. Results from this supplemental test are acceptable provided the difference between the supplemental test data and the Type A test data is within 0.25 L If results are not within 0.25 L (or 0.25 L the 1 r$as(or 0.25 L ).onshallbedetermined,correctiveactiontakes,andasuebe)s,sful supplemental test performed". j L is defined as the maximum allowable leakage rate at pressure P as shecifiedforpreoperationaltestsintheTechnicalSpecificationsor ,
associated bases, and as specified for periodic tests in the operating license.
Docket No. 50-346 License No. NPF-3 Serial No. 1436 Attachment 3 Page 4 P is defined as the calculated peak accident containment internal pSessure related to the' design basis accident and specified either in the Technical Specifications or associated bases (38 psig).
L is defined as the maximum allowable leakage rate at pressure P derived from the preoperational test data.
I't is defined as the containment vessel reduced test pressure selected to measure the integrated leakage rate during periodic. Type A tests.
ANSI N45.4-1972, Appendix C, outlines a method for verification of leakage rate tests. The method involves the accurate measurement of the leakage rate through a calibrated Jeak intentionally superimposed on the existing leaks in a containment structure.
An arrangement for superimposing a controlled and measurable leak on the containment vessel employs the orifice leak of a microadjustable instru-ment flow valve installed at a convenient penetration of the containment vessel. The flow through the valve is measured by means of a suitable flow meter. The leak orifice is selected to provide a flow under the test pressure conditions appror.imately equivalent to the leakage rate specified ;
for the containment vessel. l I
ANSI /ANS 56.8-1981 provides further guidance on performance of the supple- l mental test. It states that the leak orifice is selected to provide a j flow under the test-pressure condition equivalent to 75 percent to 125 l percent of the leakage rate specified for the containment system (L,). j During performance of the Type A leakage test, if the leakage rate is determined to be zero, a valid supplemental test could not be performed. l Therefore, since no gas would be injected into the containment or bled j from the containment during the supplemental test, the accuracy of the Type A test could not be verified. However, if the guidance of ANSI /ANS 56.8-1981 is followed regarding the 75 to 125 percent of the total allowable leakage rate for the containment being bled from the containment during the supplemental test, the accuracy of the Type A test can be verified during zero leakage conditions.
Revising Technical Specification 4.6.1.2c.3 makes the Technical Specific-ations consistent with the requirements of ANSI /ANS 56.8-1981, ANSI N45.4-1972 and current NRC position, and is identical to the License Anendment issued to the Palo Verde Nuclear Generating Station - Unit No. 1. This proposed change does not affect plant safety and will enhance the accept-ability of the Type A test. This change will also resolve NRC Inspection Report Open Iten 50-346/84029-02.
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L Docket No. 50-346
. License No. NPF-3.
Serial No. 1436 Attachment 3 L Page 5 SIGNIFICANT llA?)RDS CONSIDERATION The proposed change does not involve a significant hazards consideration because the operation of-the Davis-Besse Nuclear Power Station, Unit No. 1,'in accordance with this change would: l
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- 1. Not involve a significant increase in the probability or consequences '
of an accident previously evaluated because the proposed Technical Specification change only modifies the supplemental test used to verify the Type A test and does not modify or create any accident initiating condition (10CFR50.92(c)(1)) .
- 2. Not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed Technical Specification change only modifies the supplemental test used to verify the. Type A test and does not modify or create any accident.
initiating condition (10CFR50.92(c)(2)).
- 3. Not involve a significant reduction in a margin of safety because all USAR assumptions remain unchanged (10CFR50.92(c)(3)).
CONCLUSION Therefore, it is concluded that the proposed Technical Specification change does not involve a significant hazards consideration.
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