ML20215F823
| ML20215F823 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 10/10/1986 |
| From: | Bradley E PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20215F827 | List: |
| References | |
| NUDOCS 8610160389 | |
| Download: ML20215F823 (22) | |
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4 PHILADELPHIA ELECTRIC COMPANY 23O1 M ARKET STREET P.O. BOX 8699 PHILADELPHIA. PA.19101 COW A RD G. S A U ER. J R.
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- s".:';,L,.. 6 CUG ENE J. BR ADLEY assocente emmena6 cousses6 DON ALD SLA NMEN RUDOLPH A. CHILLEMI e.C.nIRuMALL October 10, 1986 T. H. M AHER CO RN ELL PAUL AUERB ACH messerant sessemas coumest f DW A RD J. CULLEN, J R.
THOM AS H. MILLE R, J R.
.. e....... o..s. 6 Mr. liarold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Re: Peach Bottom Atomic Power Station Units 2 and 3 Docket Nos. 50-277 and 50-278
Dear Mr. Denton:
Enclosed for filing with the Commission are 3 originals and 19 copies of a Second Amendment to Philadelphia Electric Company's Application for Amendment of Facility 0)erating Licenses DPR-44 and DPR-56 which was originally filed with tle Commission on November 18, 1976 and previously amended on April 19, 1984 This Second Amendment further revises the Application to reflect discussions with the NRC Staff and a request from the Staff to make certain clarifying changes regarding the containment leakage testing program at Peach Bottom.
Very truly yours, 7
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Euge e J. Bradley EJB:pkc j
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8610160389 861010' PDR ADOCK 03000277 3p P
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c BEFORE THE A
UNITED STATES NUCLEAR REGULATORY COMMISSION
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In the Matter of Docket Nos. 50-277 PHILADELPHIA ELECTRIC COMPANY 50-278 SECOND AMENDMENT TO NOVEMBER 18, 1976 APPLICATION FOR AMENDMENT OF FACILITY OPERATING LICENSES DPR-44 & DPR-56 Edward G. Bauer, Jr.
Eugene J. Bradley 2301 Market Street Philadelphia, Pennsylvania 19101 Attorneys for Philadelphia Electric Company i
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e BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of Docket Nos. 50-277 PHILADELPHIA ELECTRIC COMPANY 50-278 SECOND AMENDMENT TO NOVEMBER 18, 1976 APPLICATION FOR AMENDMENT OF FACILITY OPERATING LICENSES DPR-44 & DPR-56 By letter dated November 18, 1976 (E. J. Bradley (PECo) to B. C. Rusche (NRC)), Philadelphia Electric Company submitted an Application for Amendment of Facility Operating Licenses DPR-44 and DPR-56 to bring certain areas of the containment leakage testing program at Peach Bottom, as specified in Sections 3.7.A and 4.7.A of the Technical Specifications, into conformance with the requirements of 10 CFR 50, Appendix J and to request exemptions to certain provisions in 10 CFR 50, Appendix J.
By letter dated April 19, 1984 (E. J. Bradley (PECo) to H. R. Denton (NRC)), Philadelphia Electric Company submitted an Amendment to the November 18, 1976 Application For Amendment of Facility Operating Licenses DPR-44 and DPR-56 to reflect further revisions requested by the NRC Staff.
On September 10, 1985, the NRC staff requested, by telephone, that Philadelphia Electric Company submit a further amendment to its Application for Amendment for the purpose of clarifying certain items.
Accordingly, Philadelphia Electric Company, Licensee under Facility Operating Licenses DPR-44 and DPR-56 for Peach Bottom Atomic Power Station Unit No. 2 and Unit No. 3, respectively, hereby further amends its November 18, 1976 Application for Amendment, as previously amended on April 19, 1984, by deleting proposed Technical Specification pages 166, 167, 168, 168a, 169, 184, 184a, 185, 186, 187, 187a, 188, and 192 and new pages 188a and 192a referred to in the April 19, 1984 Amendment and substituting therefor updated pages 166, 167, 168, 168a, 169, 184, 184a, 185, 186, 187, 187a, 188, and 192 and new pages 187b, 188a, 188b, 188c, and 192a which are attached hereto and incorporated herein by reference.
Revised page 170 of the proposed Technical Specifications incorporated in the April 19,.., - ___ _
o 1984 Amendment Application remains unchanged and is attached hereto as a matter of completeness and is incorporated herein by reference.
The proposed changes to the existing Technical Specifications attached hereto reflect the Licensee's current request and include previously requested changes to the extent they remain applicable.
All changes to the existing Technical Specifications are indicated by a single vertical bar in the margin of the attached pages.
A discussion of the changes proposed by this Application, as amended, is set forth below.
1)
Section 4.7.A.2 of the Technical Specifications sets forth the surveillance requirements for primary containment and in particular, concerns the Integrated Leak Rate Testing program at Peach Bottom Atomic Power Station.
The value for the maximum allowable leakage rate (La) is defined as 0.5 percent of the primary containment volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 49.1 psig in Section 4.6.A.2.a on page 166 of the Technical Specifications.
The Licensee proposes, for the purpose of clarity, to delete the statement "(La is 0.5 percent)" from revised page 166 of the proposed Technical Specifications and to include the definition for the term La in the areas with the definitions for the other terms identified in the _ _ _ -., _..
formulas for the maximum allowable test leakage rate (Lt) as shown on revised pages 166 and 167 of the proposed Technical Specifications.
Additionally, the Licensee proposes to change the "0.5" in both formulas for the maximum allowable test leakage rate (Lt) to "La" as shown on revised pages 166 and 167 of the proposed Technical Specifications to be consistent with the formulas as stated in 10 CFR 50, Appendix J, Section III.A.4(a)(1)(lii).
2)
Section 4.7.A.2.e of the Technical Specifications as proposed on pages 168 and 168a of the Amendment Application, as amended on April 19, 1984, discussed requirements for determining the ILRT results for those cases when excessive leakage is identified during the performance of an ILRT and the acceptance criteria as specified in Section 4.7.A.2.d of the Technical Specifications cannot be met due to this leakage.
The Licensee proposes to revise the requirements set forth on pages 168 and 168a in order to clarify the requirements and to more explicitly state the various options available when such leakage is identified as shown on revised pages 168 and 168a of the proposed f
Technical Specifications.
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Additionally, the Licensee proposes to revise the Bases for Sections 3.7.A and 4.7.A as shown on revised page 192 of the proposed Technical Specifications to provide the basis for correcting the measured containment integrated leakage rate with local leakage measurements as specified in Section 4.7.A.2.e of the proposed Technical Specifications.
3)
Section 4.7.A.2.f of the Technical Specifications as shown on page 169 discusses the requirements for performing Local Leak Rate Tests (LLRTs) on primary containment testable penetrations and isolation valves.
The Licensee proposes to revise these requirements to include reference to Technical Specification Tables 3.7.2, 3.7.3, and 3.7.4 as shown on revised page 169 of the proposed Technical Specifications.
These tables list the primary containment isolation valves and the testable penetrations to which Section 4.7.A.2.f applies and references the applicable notes which provide detailed information concerning their testing methods.
4)
Table 3.7.2, as shown on page 184 of the proposed Technical Specifications, lists the testable penetrations with double O-ring seals.
The Licensee proposes to revise the nomenclature for Penetration No.
N-2 to correctly identify it as the " Equipment Access.
and Personnel Air Lock" penetration.
Additionally, the Licensee proposes to re-insert Note (7) on revised page 188 and as referenced on revised page 184 of the proposed Technical Specifications for Penetration No. N-2 to identify the testing frequency for the personnel air locks.
Personnel air lock testing is consistent with the provisions of 10 CFR 50, Appendix J, Section III.D.2(b).
1 The Licensee also proposes to correct a typographical error by deleting the quotation mark (") after Stop J
Check 23-12 on Penetration No. N-214 as shown on revised page 184 of the proposed Technical Specifications.
5)
Table 3.7.3, as shown on page 184a of the proposed Technical Specifications, lists the testable penetrations with testable bellows.
The Licensee i
J proposes to revise Table 3.7.3, as shown on revised page 184a of the proposed Technical Specifications, to correct the nomenclature for particular penetrations to reflect the present plant nomenclature for those penetrations.
t 6)
Table 3.7.4 of the proposed Technical Specifications, submitted in the April 19, 1984, Amendment, listed
" testable" containment isolation valves at Peach Bottom l
Units 2 and 3 along with comments regarding the testing f :
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of these valves.
Subsequently, the NRC staff requested that all. primary containment isolation valves, including those valves which are exempt from testing under the requirements of 10 CFR 50, Appendix J, be listed in Table 3.7.4.
Accordingly, the Licensee proposes for the purpose of clarity to replace Table 3.7.4 in its entirety as presented on pages 185, 186, 187, 187a, 188 and new page 188a of the April 19, 1984 Amendment with the revised Table 3.7.4 as shown on revised pages 185, 186, 187, 187a, 188 and new pages 187b, 188a, 188b and 188c of the proposed Technical Specifications.
The footnotes as shown on page 185 of the proposed Table 3.7.4 were revised to reflect the present plant configuration as the result of modifications performed subsequent to the April 19, 1984 Amendment.
The proposed Table 3.7.4 reflects the addition of testable isolation valves on penetration numbers 234, and 234B as the result of the installation of a new Post-Accident Sampling System (PASS).
Penetration No. 14 on page 185 of the proposed Technical Specifications is being revised to differentiate between the MO-12-15 valves on Unit Nos. 2 and 3 along with the applicable reference notes since the MO-12-15 valve on
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. Unit 2.11s a gate valve and the MO-12-15 valve on Unit 3 is a globe valve.
Penetration 35F on page 186 of the proposed Technical Specifications is being revised to correctly identify SV-109 as the testable isolation valve for that penetration.
Note (9) on page 188 of the proposed Technical Specifications is being revised and included as a 4
reference on revised page 185 of the proposed Technical Specifications for Penetration Nos. 7A to D to indicate l
that the Main Steam Isolation Valves may be tested by 1
pressurizing between the inboard and outboard valves.
Note (10) on revised page 188 of the proposed Technical Specifications is being revised to include a statement that the identified test method is an exemption to 10 CFR 50, Appendix J and to indicate that the MO-12-15 valve on Unit No. 2 is a gate valve tested in the reverse direction as referenced on revised page 185 of the proposed Technical Specifications.
Note (11) on new page 188a of the proposed Technical Specifications is being revised to identify the inboard Main Steam Isolation Valves as globe valves which may be tested in the reverse direction as referenced on page 3
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185.of the proposed Technical Specifications and to indicate that the MO-12-15 valve on Unit No. 3 is a globe valve tested in the reverse direction as referenced on revised page 185 of the proposed Technical Specifications.
The Bases for Sections 3.7.A and 4.7.A are being revised as shown on revised page 192a of the proposed Technical Specifications to indicate that testing of the inboard globe valves listed in Note (11) on page 188a of the proposed Technical Specifications in the reverse direction, yields conservative results since the globe is lifted off its seat.
Note (12) is being revised as shown on new page 188a of the proposed Technical Specifications to delete valves AO-2502B and AO-3502B from the note since these valves were replaced with different type valves and Note (12) is no longer applicable.
A new Note (22), which is a
applicable to these valves, is being added to new page 188c of the proposed Technical Specifications.
Note (12) is being changed to Note (22) as a reference on revised page 187a of the persposed Technical Specifications for Penetration No. 205A to reflect this change.
Additionally, a reference to Note 12 for Penetration No. 25 is being added as shown on revised -
page 186 of the proposed Technical Specifications to indicate that valves AO-2502A and AO-3502A are butterfly valves tested in the reverse direction.
Note (13) is being revised as shown on new page 188a of the proposed Technical Specifications to include appropriate supplemental information concerning the testing of the identified valves.
Note (19) is being added to new page 188c and referenced on revised page 186 of the proposed Technical Specification for Penetration Nos. 35A to E to indicate that the Transversing In-core Probe (TIP) shear valves are not Type C tested since squib detonation is required for closure.
Note (20) is being added to new page 188c and referenced on revised page 187 of the proposed Technical Specifications for Penetration No. 42 to indicate that the referenced valves are explosion valves tested in the reverse direction.
Note (21) is being added to new page 188c and referenced on revised page 187 of the proposed Technical Specifications for Penetration No. 102B to indicate that the Breathing Air System valves on Unit No. 3 are manual gate valves tested in the reverse direction..
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7)
The Licensee, in its April 19, 1984 Amendment,~ deleted the following gate valves from Table 3.7.4:
MO-14-70,
" Torus Water Clean-Up Pump Suction from Suppression Pool", MO-23-58, "High Pressure Coolant Injection Pump Suction from Suppression Pool", and MO-13-41, "Resctor Core Isolation Cooling Pump Suction from Suppression Pool".
The justification for the deletion was that these valves are water covered throughout the post-accident period and are not relied upon to prevent the escape of containment atmosphere.
On September 10, 1985, the NRC staff, in a telecon, requested that the valves remain subject to periodic leak rate testing; however, testing in the reverse direction was found to be acceptable.
Accordingly, the Licensee, proposes to include gate valves MO-14-70, MO-23-58, and MO-13-41 as testable isolation valves on revised Table 3.7.4 as indicated by Note (16) on new page 187b of the proposed Technical Specifications.
Note (16) is being added to new page i
188b of the proposed Technical Specifications to indicate that these valves are gate valves tested in the i
reverse direction.
8)
The Licensee, by letter dated May 15, 1981 from S.
L.
Daltroff (PECo) to J. F. Stolz (NRC), requested an 5,
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exemption from Local Leak Rate Testing of a group of water-covered torus isolation valves which are located in turbine exhaust lines from the High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) turbines, in pump suction lines to the Core Spray pumps.and in various minimum flow recirculation lines or other recirculation / vent lines from the Residual Heat Removal (RHR), HPCI, RCIC and Core Spray Systems.
The.
NRC staff informed the Licensee that since these isolation valves are water sealed throughout the post-accident period by suppression pool water, they would not require local leak rate testing under the requirements of Appendix J.
Subsequently, the Licensee deleted these valves from Table 3.7.4 as set forth in the April 19, 1984 Amendment.
Although exempt from Appendix J testing, the NRC staff requested in the September 10, 1985 telecon that these valves continue to be listed on Table 3.7.4.
An inspection of the Peach Bottom Unit 3 water covered discharge lines in the suppression chamber identified four lines with anti-syphon devices that permit the lines to communicate directly with the gas space above the suppression pool.
They involve the RHR minimum flow check valves10-19A,C (penetration 210B) and 10-19B,D (penetration 210A), RCIC Vacuum Pump discharge stop -
check valve 13-10 (penetration 221), and HPCI Turbine exhaust drain stop check valve 23-13 (penetration 223).
A modification will be performed either during the Unit 3 mid-cycle examination outage currently scheduled for January, 1987, or during the next Unit 3 refueling outage, whichever comes first, to seal the anti-syphon devices to preclude communication with the suppression pool air space.
An inspection of the Peach Bottom Unit 2 water covered discharge lines in the suppression chamber will be conducted during the next Refueling Outage scheduled to begin March, 1987, and modifications will be performed as necessary during that time to preclude communication with the suppression pool air space.
A footnote is being added to Table 3.7.4 on revised page 187a of the proposed Technical Specifications to indicate which notes become effective following completion of these modifications.
Accordingly, the Licensee proposes to include those torus isolation valves identified under Exemption D of the Licensee's May 15, 1981 letter on revised Table 3.7.4 as indicated by Note (15) or Note (17) on revised page 187a and new page 187b of the proposed Technical Specifications.
Note (15) is being added to new page 188b and referenced on revised page 187a of the proposed Technical Specifications for Penetration No. 217B and new page 187b of the proposed Technical Specifications for Penetration Nos. 221 and 223 to indicate that the referenced stop-check valves serve only as block valves and that the check function of these valves is not leak tested.
Note (17) is being added to new pages 188b and 188c and referenced on revised page 187a and new page 187b of the proposed Technical Specifications to indicate that the applicable valves are tested in conformance with the requirements of ASME,Section XI, in lieu of Appendix J, Type C, leak rate testing.
9)
The Licensee, in its May 15, 1981 letter, requested an exemption from local leak rate testing of the individual isolation valves in the Control Rod Drive (CRD) insert and withdraw lines to the CRD hydraulic control units.
The NRC staff informed the Licensee that Appendix J does not require local leak rate testing of these valves since the insert and withdraw lines represent closed systems with regard to containment atmosphere and are constantly under water pressure from reactor vessel liquid level at reactor vessel pressure.
Subsequently, the April 19, 1984 Amendment deleted these valves from Table 3.7.4.
Although exempt from Appendix J testing, the NRC staff requested in the September 10, 1985 telecon that all containment isolation valves be listed on Table 3.7.4. -.
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Accordingly, the Licensee proposes to include the isolation valves in the CRD insert and withdraw lines to the hydraulic control units on revised Table 3.7.4 as identified by Hote (18) on revised page 186 of the proposed Technical Specifications.
Note (18) la being added to new page 188c and referenced on revised page 186 of the proposed Technical Specifications to indicate that individual valves on the CRD hydraulic control units are not Type C tested.
Safety Analysis The proposed changes revise the containment leakage testing program in certain areas to conform with the requirements of 10 CFR 50, Appendix J or the NRC requests provided in the September 10, 1985 telecon.
Exemptions to the testing provisions of 10 CFR 50, Appendix J, were previously identified in the Licensee's August 13, 1976, November 17, 1976, and May 15, 1981 letters.
Those exemptions deemed acceptable to the NRC, based on previous discussions on this issue, are reflected in this application.
The valve design and testing features that establish the acceptability for these exemptions are stated in the notes for Tables 3.7.2 through 3.7.4 of the Technical Specifications. -
The containment testing requirements of the Technical Specifications provide for periodic verification of the leak tight integrity of the primary reactor containment, and systems and components which penetrate containment, and establish the acceptance criteria for such tests.
In this manner, the release of radioactive materials from the containment atmosphere will be restricted to the leak rates assumed in the safety analyses.
This restriction will limit off-site radiation doses to a small fraction of the Commission's regulations during accident conditions.
The changes proposed in the November 18, 1976 Application, as amended on April 19, 1984 and herein, establish additional conservatism in the performance of containment testing, most notably by the increase in the isolation valves subject to the local leak rate requirements of the Technical Specifications, and the more conservative criteria for computing the integrated leak rate for the containment.
Significant Hazards Consideration Determination The Commission has provided guidance for determining whether a significant hazards consideration exists by providing examples of amendments that are considered not likely to involve significant hazards consideration (48 FR 14870).
One such example of an action involving no significant hazards consideration is a change that constitutes an additional limitation, restriction, or control not presently included in the --
Technical Specification.
The November 18, 1976 License Amendment application, as amended on April 19, 1984 and herein, meets this example.
These changes do not involve a significant hazards consideration since they do not:
(1) involve a significant increase in the probability or consequences of an accident previously evaluated since the changes enhance the conservatism of both the integrated and local leak rate testing program; (2) create the possibility of a new or different kind of accident from any accident previously evaluated since additional surveillance provisions do not create a potential accident precursor; (3) involve a significant reduction in a margin of safety since the addition of isolation valves to the testing program and the more conservative leak rate criteria proposed by the Application improves the assurance that containment integrity will be maintained, and off-site radiation doses will be limited to a small fraction of the Commission's regulations during accident conditions.
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The Plant Operation Review Committee and the Nuclear Review Board have reviewed the proposed changes to the Technical Specifications and have concluded that they do not involve an unreviewed safety question or a significant hazards consideration and will not endanger the health and safety of the public.
Respectfully submitted, PHILADELPHIA ELECTRIC COMPANY l
By
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fice Pc'esideht N
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COMMONWEALTH OF PENNSYLVANIA ss.
COUNTY OF PHILADELPHIA J. S. Kemper, being first duly sworn, deposes and says:
That he is Vice President of Philadelphia Electric Company, the Applicant herein; that he has read the foregoing Second Amendment to Application for Amendment of Facility Operating Licenses and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief.
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f Subscribed and sworn to
/h-before me this /d day of &
ly lffh
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yQ Md Notary Public MELANIE R. CAMPANELLA Notary Public, Philadelphia, Phdade!phia Co.
My Commission Espires February 12.1%0
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l CERTIFICATE OF SERVICE i
I certify that service of the foregoing Second Amendment was made upon the Commonwealth of Pennsylvania, by mailing a copy thereof, via first-class mail, to Thomas R. Gerusky, Director, Bureau of Radiological Protection, P. O.
Box 2063, liarrisburg, PA 17120; all this loth day of October, 1986.
T l
E ene J. Bradley Attorney for Philadelphia Electric Company W
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