ML20244D989
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JUN 1/1986 MEMORANDUM FOR: Victor Nerses, Project Manager PWR Project Directorate #5 Division of PWR Licensing-A FROM:
Charles E. Rossi, Assistant Director Division of PWR Licensing-A
SUBJECT:
INPUT FOR SUPPLEMENT TO THE SAFETY EVALUATION REPORT FOR THE SEABROOK PLANT, UNITS 1 AND 2 Plant Name: Seabrook Plant, Units 1 and 2 Docket Nos.: 50-443/444 Licensing Stage: OL NSSS Supplier: Westinghouse Architect Engineer: United Engineers and Constructors Containment Type:
Dry Responsible Branch: PWR Project Directorate #5 Review Status: Complete The enclosed input for a Supplement to the Safety Evaluation Report (SSER) for the Seabrook Plant, Units 1 and 2 (Enclosure 1) has been prepared by the Engineering Branch after having reviewed the applicant's letters dated May 7 and 21, 1986, concerning the following:
1.
Request for exemption from requirements of Appendix J to 10 CFR 50 regarding containment air lock testing.
2.
Revised Leakage Rate Testing Program for Containment Isolation Valves (Type C Testing), concerning applicant proposals made since the publication of the SER.
In the enclosed SSER input, we find the applicant's proposals regarding these issues to be acceptable, and consider these issues to be resolved.
A brief SALP report is enclosed.
sf g bDpz. L( oO}C] 1(tf3(P isinsi signe5 by Charles E. Rossi, Assistant Director Division of PWR Licensing-A
Enclosure:
As stated cc:
V. Noonan DISTRIBUTION:
C. Early-Docket Files R. Ballard PAEB Reading Files G. Bagchi PAEB Plant Files F. Rinaldi J. Pulsipher
Contact:
J. Pulsipher X-27793 See Previous Concurrence Sheet
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MEMORANDUM FOR:
Victor Nerses, Project Manager
\\
PWR Project Directorate #5 Division of PWR Licensing-A M:
Charles E. Rossi, Assistant Director Division of PWR Licensing-A SUBJE :
INPUT FOR SUPPLEMENT TO THE SAFETY EVALUATION REPORT FOR THE SEABROOK PLANT, UNITS 1 AND 2 Plant Name: Seabrook Plant, Units I and 2 Docket Nos.:
50-443/444 Licensing Stag - OL NSSS Supplier:
westinghouse Architect Enginee - United Engineers and Constructors j
Containment Type:
ry Responsible Branch:
WR Project Directorate #5 Review Status: Contin ing l
The enclosed input for a pplement to the Safety Evaluation Report (SSER) l for the Seabrook Plant, Un s I and 2 (Enclosure 1) has been prepared by l
the Engineering Branch after aving reviewed the applicant's letters dated May 7 and 21,1986, concernin the following:
1.
Request for exemption om requirements of Appendix J to 10 CFR 50 regarding containment al lock testing.
l 2.
Revised Leakage Rate Testi Program for Containment Isolation Valves (Type C Testing), concerning applicant proposals made since the publication of the SER.
In the enclosed SSER input, we find the ap licant's proposals regarding these issues to be acceptable, and consider hese issues to be resolved.
A brief SALP report is enclosed.
Charles E. Rossi, Ass tant Director Division of PWR Licensi
-A
Enclosure:
As stated cc:
V. Noonan DISTRIBUTION:
C. Early Docket Files R. Ballard PAEB Reading Files G. Bagchi PAEB Plant Files j
F. Rinaldi i
J. Pulsipher 1
Contact:
J. Pulsipher l
X-27793 See Previous Concurrence Sheet
- 1 I
PAEB*
PAEB*
PAE PWR:A:A i
JPulsipher:vt GBagchi RLBal rd CERossi I
6/
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Victor Nerses We are continuing to review the following issues, concerning containment ystems topics:
1.
Reanalysis of containment pressure and temperature transient during LOCA, contained in applicant letter dated May 17, 1986.
We will ovide our input concerning these issues in ar upcoming input to the SSER.
A brief SALP port is enclosed.
l Charles E. Rossi, Assistant Director Division of PWR Licensing-A
Enclosure:
As.4tated cc:
V. Noonan C. Early R. Ballard G. Bagchi F. Rinaldi J. Pulsipher
Contact:
J. Pulsipher X-27793 DISTRIBUTION:
DMB Docket Files PAEB Reading Files PAEB Plant Files CERossi PA d
PAEB R:A:AD JP ef:vt GBaJchi RLBallard CE ssi 6/
/86 6/ l
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a e
1 6.2.6 Containment Leakage Testing l
Containment Air Lock Surveillance By letter dated May 21, 1986, the applicant requested an exemption from certain requirements of Appendix J to 10 CFR 50. The staff's evaluation of this request for exemption follows:
Paragraph III.D.2(b)(ii) of Appendix J states:
" Air locks opened during periods when containment integrity is not required by the plant's Technical Specifications shall be tested at the end of such periods at not less than Pa."
Whenever the plant is in cold shutdown (Mode 5) or refueling (Mode 6),
7 containment integrity is not required. However, if an air lock is opened during Modes 5 and 6, paragraph III.D.2(b)(ii) of Appendix J requires that an overall air lock leakage test at not less than Pa be conducted prior to plant heatup and startup (i.e., entering Mode 4). The existing air lock doors are so designed that a full pressure [i.e., Pa (49.6 psig)) test of an entire air lock can only be perfomed after strong backs (structural bracing) have been installed on the inner door. Strong backs are needed since the pressure exerted on the inner door during the test is in a direction opposite to that of the accident pressure direction. Installing strong backs, performing the test, and removing the strong backs, requires several hours, during which access through the air lock is prohibited.
If the periodic 6-month test of paragraph III.D.2(b)(i) of Appendix J and the test required by paragraph III.D.2(b)(iii) of Appendix J are current, no maintenance has been perfonned on the air lock that could affect its sealing capability, and the air lock is properly sealed, there should be no reason to expect the air lock to leak excessively just because it has been opened in Mode 5 or Mode 6.
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a ENCLOSURE 1 SEABROOK SSER - J. PULSIPMER l
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y i Accordingly, the staff concludes that the applicant's proposed approach of substituting the seal leakage test of paragraph III.D.'2(b)(iii) for the full pressure test of paragraph III.D.2(b)(iil of Appendix J is acceptable when no maintenance that could affect sealing capability has been performed on an air lock. Whenever maintenance that could affect sealing capability has been perfonned on an air lock, the requirements of paragraph III.D.2(b)(ii) of Appendix J must still be met by the licensee.
The'special circumstances for granting this exemption pursuant to 10 CFR 50.12 have also been identified. The purpose of Appendix J to 10 CFR 50 is to assure that containment leak-tight integrity can be verified periodically throughout service lifetime so as to maintain containment I
leakage within the limits specified in the facility Technical Specifications.
The proposed alternative test method is sufficient to achieve this underlying purpose in that it provides adequate assurance of continued leak-tight integrity of the airlock.
In addition, at the time this section of Appendix J was revised in 1980, the staff did not contemplate the undue hardship and cost which would result from the requirement to perform a time-consuming full-pressure test before starting up from eye.., the shortest cold shutdown during which the airlock had been used for containment entry.
Because of this, the staff has already granted this same exemption to numerous plants, and intends to revise Appendix J to alleviate the need for further similar exemptions.
Consequently, the special circumstances described by 10 CFR 50.12(a)(2)(ii) and (iii) exist' in that application of the regulation in these particular circumstances is not necessary to achieve the underlying purpose of the rule in that the applicant has proposed an acceptable alternative test method that accomplishes the intent of the i
regulation. Compliance would result in undue hardship that is significantly in excess of that contemplated when the regulation was
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adopted and that is significantly in excess of that incurred by others similarly situated in that plant startup is delayed and unnecessary personnel radiation exposures are incurred while performing an overall airlock leakage test at full pressure.
O Therefore, a partial exemption from' this requiremer,t [10 CFR 50, Appendix J, paragraph III.D.2(b)(ii)] is justified and acceptable for Seabrook, Units 1 and 2; and the applicant's proposal to adopt surveillance requirement 4.6.1.3 of Revision 4 of NUREG-0452, " Standard Technical Specifications for Westinghouse Pressurized Water Reactors", is acceptable (NUREG-0452, Rev.
4, is written to accommodate this exemption).
Further, the staff finds that, in accordance with 10 CFR 50.12(a)(1), the requested partial exemption is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security; and that, in accordance with 10 CFR 50.12(a)(2)(ii) and (iii), special circumstances, as discussed above, are present.
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j Type C Testing Program The SER contains the staff's review and acceptance of the applicant's program for local leakage rate testing of containment isolation valves, designated as Type C testing in Appendix J to 10 CFR 50. However, the applicant has proposed, by letter dated May 7,1986, to revise their Type C testing program. The following is a discussion of the penetrations and valves that will not be tested under the revised program; the remaining cc ainment isolation valves will be Type C tested.
Containment penetrations X-1, X-2, X-3, X-4, X-5, X-6, X-7, X-8, X-48A, X-488, X-49A, X-49B, X-63, X-64, X-65, and X-66 serve the Main Steam,
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Feedwater, Primary Component Cooling Water, and Steam Generator Blowdown l
Systems. The applicant does not propose te Type C test the valves associated with these secondary system penetrations.
This is not a change in the testing prcgram; the original program contained the same provision, and was specifically reviewed and found acceptable by the staff in the Seabrook SER. The staff coatinues to find this acceptable. To briefly restate the basis for the SER conclusion, the secondary system is a closed system inside containment, not p,0stulated to rupture during an accident; it will preclude containment atmos'phere from reaching the associeted isolation valves, and so the valves will not be relied on to limit containment leakage. Although the Primary Component Cooling Water system is not part of the secondary system, it also fonns a closed system inside containment and this argument applies equally to it. Therefore, the valves in these l
lines are not required to be Type C tested.
The integrity of these closed systems inside containment will be tested during preoperational and periodic Type A (Integrated Leakage Rate) testing.
Containment penetrations.Y-71A/74A, X-71B/74B, X-72A/75A, and X-72B/75B are the supply and returr lines for the two hydrogen gas analyzers (CGC-CP-173
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I andCGC-CP-174).
These lines form two closed systems (Train 'A' and Train
'B') outside containment which satisfy the following design criteria.
The systems are Seismic Category I.
The systems are Safety Class 2 up to the analyzer isolation valves downstream of the containment isolation valves.
The isolation valves in these lines are expected to be open throughout an accident, and the closed systems outside containment will become extensions of the containment boundary.
Standard Review Plan (SRP) 6.2.6, " Containment Leakage Testing, in Section II, states:
Leak testing of instrumentation lines that penetrate containment may be done in conjunct 4n with either the local leak rate tests or the containment integrated leak rate test.
Instrumentation lines that are not locally leak tested should not be isolated from the containment athiosphere during the performance of the CILRT.
These isolation valves, therefore, will not be Type C tested. The integrity of the closed systems outside containment will be tested by epening the isolation valves during preoperational and periodic Type A (CILRT) testing, thereby subjecting each system boundary to postulated post-accident containment pressure.
In accordance with SRP 6.2.6, the staff finds this to be acceptable.
Several containment penetrations associated with the Emergency Core Cooling Systems (ECCS) will realize a water seal maintained at a pressure grcater than 1.10 Pa following a LOCA. The ECCS satisfies the following design criteria.
. The system is Safety Class 2.
The system is Seismic Category I.
The system is protected against missiles.
The system is protected against the dynamic effects associated with pipe ruptures.
1 Active electrical components are classified as IE and receive emergency power from the diesel generators.
Failure of a diesel generator results in the loss of one train of active ECCS components.
The redundant diesel will continue to power the fully redundant ECCS Train.
i The penetrations at which this pressurized water seal will be maintained throughout the entire 30 day accident mitigation are X-24, X-25, X-26, X-27, X-28, X-29, X-30, X-31, and X-33.
In accordance with paragraph III.C.3. ef Appendix J to 10 CFR 50, these valves ara not required to be Type C tested, as they are, in effect, sealed with water from seal systems with unlimited supplies of sealing water from the containment sump.
Details of the hydraulic conditions at the penetrations and related system operation are described below.
Some of the containment isolation valves involved are expected to remain open throughout the accident, in which case they would not be performing a containment isolation function and valve j
leakage would be irrelevant; however, if they were to close for any reason (e.g., valve failure), the pressurized water seal would be maintained and would prevent the valves from becoming potential containment atmosphere leak paths.
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. The post-accident operation of the ECCS can be described in 3 phases.
Immediately following the accident, the RHR pumps, the ce.itrifugal charging pumps, and the safety injection (SI) pumps are all aligned to take suction from the Refueling Water Storage Tank (RWST) and discharge to all' four Reactor Coolant System (RCS) cold legs (cold leg injection).
In the presence of the 'P' signal (Phase 'B' isolation), the containment spray (CS) pumps also take suction from the RWST and discharge to the spray ring headers.
When the RWST reaches the " low-low" level, the containment sumps' isolation valves are manually opened and the system enters the. cold leg recirculation mode. During this phase, the RHR pumps and containment spray pumps take suction from the containment sumps. The RHR pumps continue discharging to the RCS cold legs, and also deliver water to a suction header common to the safety injection and charging pumps. These pumps also continue. discharging to the RCS cold legs. Approximately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> after the onset of an accident, the system ic Switched over to the hot leg recirculation mode. During this phase, the RHR and containment spray pumps continue taking suction from the containment sumps. The RHR pumps are aligned to discharge to RCS hot legs 1 and 4, as well as the SI/CS pumps' comon suction header. The safety injection pumps are' aligned to discharge to all four RCS hot legs, and the charging pumps continue discharging to all four RCS cold legs.
Penetration X-24 is the common discharge line from the centrifugal charging pumps to all four RCS cold legs. The isolation valves outside containment for the penetrai; ion (SI-V138 and SI-V139) will be open for the entire accident mitigation process. Both charging pumps discharge to a comon header throughout accident mitigatian. Therefore,' a water seal at a pressure greater than 1.10 Pa will be maintained at the penetration for 30 days following the onset of an accident regardless of single active failure.
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h l Penetrations X-25 and X-26 are the discharge lines frce the safety injection pumps to all four RCS hot legs.
The isolation valves outside containment for these penetrations (SI-V77 and SI-V102) will be closed for cold leg injection and cold leg recirculation and open during hot leg recirculation. During cold leg injection and recirculation, the crosstie line between both pumps' discharge lines will not be isolated. 1herefore, regcrdless of single active failure a water seal at a pressure greater than 1.10 Pa will still be maintained at these penetrations by the redundant SI pump. During hot leg recirculation, the crosstie isolation valves (SI-VIII and SI-V112) are closed. However, during both recirculation phases the RHR pump (s) deliver water to a suction header common to both SI pumps. During preoperational testing with one RHR pump delivering to this suction header, l
the suction pressure for both SI pumps was in excess cf 1.10 Pa.
Therefore, should a single active failure occur, a water seal pressurized to 1.10 Pa will still be maintained at the penetration corresponding to the failed SI pump by the lone operating RHR pump.
It follows that a pressurized water seal will be maintained at these penetrations for 30 days following the onset of an accident regardless of single active failure.
Penetretix X-27 is the discharge line to all four RCS cold legs from the safety injection pumps. The isolation valve outside containment for this penetration (SI-V114) will be open during cold leg injection and cold leg recirculation and closed during hot leg recirculation. This line is a continuation of the crosstie line between both pumps' discharge lines.
Given the above discussion on penetrations X-25 and X-26, it follows that a I
pressurized water seal will be maintained at penetration X-27 for 30 days j
l following the onset of an accident regardless of single active failure.
Penetrations X-28, X-29, X-30: and X-31 are the centrifugal charging pumps' discharge lines to the reactor coolant pumps' seals. The isolation valves outside containment for these penetrations (CS-V154, CS-Y158, CS-V162, and CS-V166) do not close automatically following an accident and may remain i
open during accident mitigation. These lines sre supplied with water from the l
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4 discharge header common to both charging pumps.
Therefore, a water seal at a pressure greater than 1.10 Pa will be maintained at these penetrations for 30 days following the onset of an accident regardless of single active failure.
Penetration X-33 is the normal charging line from the centrifugal charging pumps' common discharge header. The isolation valve outside containment i
for this penetration (CS-V143) will be clcsed throughout the entire f
accident mitigation process. Because the line is pressurized by the discharge header common to both charging pumps, the penetration will realize a water seal at a pressure greater than 1.10 Pa for 30 days following an accident regardless of single active failure.
As previously discussed, a water seal at a pressure greater than 1.10 Pa will be maintained at penetrations X-24, X-25, X-26, X-27, X-28, X-29, X-30, X-31, and X-33 for 30 days following the onset of an accident. This seal precludes any isolation valve seat leakage of containment atmosphere.
In addition, all containment isolation valves for these penetrations located outside containment which may be closed at some time following an accident are wedge-type gate valves. The applicant states that their f
design allows for stem / packing leakage only from the high pressure side of the wedge. Given that a water seal will be maintained on the side of the wedge gate away from containment at a pressure greater than 1.10 Pa, containment atmosphere leakage from the stems or packing on these valves would therefore be precluded. Therefore, the containment isolation valves associated with the penetrations listed above are not potential containment atmosphere leak paths.
In accordance with paragraph III.C.3. of Appendix J, these valves are not subject to Type C testing.
Containment penetrations X-60 and X-61 are the suction lines for the containment spray and RHR pumps from the containment sumps. Each of these penetrations has one isolation valve located outside containment (CBS-Y8 and CBS-V14) which is normally closed and will remain closed during cold i
. 1eg injection, and will be open during cold and hot leg recirculation.
The containment recirculation sumps and these associated penetrations will fill with water almost immediately following the onset of an accident, and will terrain filled with water for at least 30 days following the onset of an accident. These valves, therefore, do not constitute potential containment atmosphere leak paths, and as such are not required by paragraph III.A.I.(d) of Appendix J to be Type C tested. Paragraph II.H. of Appendix J also defines four categories of valves that require Type C testing; they are those that:
1)
Provide a direct connection between the atmosphere inside and outside containment under normal operation; 7
2)
Are required to close automatically upon receipt of a containment isolation signal; 3)
Are required to operate intermittently under post-accident conditions; and 4)
Are in main steam and feedwater piping of boiling water reactors (BWRs).
The valves in question are opened, rather than closed, and remain open for the duration of the accident. Thus, paragraph II.H. does not require them to be tested. The staff concludes, therefore, that the valves in penetrations X-60 and X-61 are not subject to Type C testing.
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In conclusion, the staff finds, based on the above evaluation, that the applicant's proposed Type C testing program, as described in their letter dated May 7,1986, complies with the requirements of Appendix J to 10 CFR 50 and is acceptable.
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ATTACHMENT ?
SALP REPORT PLANT NAME:
Seabrock Plant APPLICANT:
Public Service Co.
DOCKET NO.:
F0 443 ar.d 50-444 REVIEWER:
.1. Pultipher LICENSING ACTIVITY:
SSFD for
-Fectinn f.P.6 l
EVAltfAT10N CRITERI A RATING Description l
1.
Planagement !nyn1venent NA No basis
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Approach to Resolution
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Resolution o' the icetw of Technical Issues from was viable and sourd a safety standpoint frem a safety standpoint.
3.
Responsiveness to NDC 2
The licensee resocedet initiatives to staff-ccreerns in 7
an adequate manner.
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Enforcement History NA 5.
Reporting ard Analysis NA c' Reporteble Ever.ts 6.
Staffing t'A 7.
Training NA
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