ML19332F730
| ML19332F730 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 12/11/1989 |
| From: | Feigenbaum T PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TASK-***, TASK-OR GL-89-21, NYN-89162, NUDOCS 8912180189 | |
| Download: ML19332F730 (19) | |
Text
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...g; ew Hampshire j
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'/ T Senior Vice President and.
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Chief Operating Officer 3
NYN-89162 December 11, 1989 1
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n United States Nuclear Regulatory Commission F
Washington, DC 20555 y
i Attention: Document Control Desk References (a)
Facility Operating License No. NPF-67. Docket No. 50-443 (b)
USNRC Generic Letter 89-21, dated October 19, 1989,
-' Request for Information Concerning Status of Implementation of. Unresolved Safety Issue (USI)
Requirements
Subject:
Response to Generic Letter 89-21 Gentlemen:
New-Hampshire Yankee provides herewith the information requested by Generic' Letter 89-21 regarding the status of implementation at Seabrook Station of Unresolved Safety Issues for which final technical resolution
-has.been achieved. to this letter is a table, structured in the format of-to Generic Letter 89-21, which has been completed as requested.
The table-indicates which Unresolved Safety Issues are not applicable to Seabrook Station.
Specific implementation status informati.on is provided for those Unresolved Safety Issues which are applicable to Seabrook Station.
If you have any questions on this matter, please contact Mr. Geoffrey Kingston at (603) 474-9521, extension 3371.
Sincerely your,
rgl/
dek Ted C. Feigen Enclosure l
I-891218o189 891211 hDR ADOCK 0500o443 kp l
PDC I \\
1 l-New Hampshire Yankee Division of Public Service Company of New Hampshire i
P.O. Box 300
- Seabrook, NH 03874
- Telephone (603) 474 9521 l
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United States Nuclear Regulatory Commission December 11, 1989 Attention: Document Control Desk Page 2 l
cci Mr. William T. Russell Regional Administrator United States Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Mr. Victor Nerses Project Manager Project Directorate I-3 United States Nuclear Regulatory Commission Division of Reactor Projects Washington, DC 20555 l
Mr. Antone C. Cerne NaC Senior Resident Inspector P.O. Box 1149 Seabrook, NH 03874 l
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December 11, 1989 ENCLOSURE 1 TO NYN-89162 UNRESOLVED SAFETY ISSUES FOR WHICH A FINAL TECHNICAL RESOLUTION HAS BEEN ACHIEVED l
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-December.Ili1989 UNRESOLVED SAFETY ISSUES FOR WHICH A S NAL TECHNICAL RESOLUTION MAS BEEN ACHIEVED USI/WA EEIEBER TITLE REr. DOC 5HEET APPLICASILITY STATES / BATE
- REINIAE5**
A-1 Water Hammer SECY 84-119 All C/06-89 A-1 NUREG 0927. Rev. 1
-NUREG-0993 Rev. 1 RUREG-0737 Item I.A.2.3 SRP Revisions A-2 Asymmetric Blowdown NUREG-0609 PWR C/02-85 A-2 MPA D-10 Loads On Reactor Primary GL-84-04, GDC-4 Coolant Systems SBN-756 SBN-703 A-3 Westinghouse Steam NUREG-0844 W-PWR I/12-89 A-3 Generator Tube Integrity SECY 86-97 SECY 88-272 GL 85-02
)
SBN-815 (No requirements) 4 1
A-4 CE Steam Generator Tube NUREG-0844. SECT 86-97 CE-PWR N/A Integrity SECY 88-272 GL 85-02 (No requirements)
A-5 B&W Steam Generator MUREG-0844, SECY 86-97 B&W-PWR N/A Tube Integrity SECY 88-272 GL 85-02 (No requirements)
E A-6 Mark I Containment MUREG-0408 Mark I-BWR N/A Short-Term Program
- templete NC - No changes necessary
- See pages (6) through (16) of this Enclosure for remerts.
BA - Not applicable I - Incomplete E - Evaluating actions required (1)
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~ December 11~.1989_
BSI/MPA BE M **
stNBER TITLE REF. DOCWEET APPLICMILITY STATUS / BATE
- A-7/
Mark I long-Term NUREG-0661 Mark I-BWR N/A D-01 Program
'N'JREG-0661, Suppl. 1 GL 79 A-8 Mark II Containment NUREG-0808-Mark II-BWR N/A Pool Dynamic Loads NUREG-0487, Suppl. 1/2 MUREG-0802 SRP 6.2.1.1C GDC 16 A-9 Anticipated Transients NUREG-0460. Vol. 4 All I/08-31-90 A-9 Without Scram 10 CFR 50.62 NYN-89084 A-10 BWR Feedwater Mozzie NUREG-0619 BWR N/A MPA B-25 Cracking Letter from DG Eisenhut dated 11/13/80 GL 81-11 A-11 Reactor Vessel Material NUREG-0744, Rev. 1 All N/C A-11 Toughness 10 CFR 50.60/.
GL 82-26 A-12 Fracture Toughress of NUREG-0577, Rev. 1 PWR N/A A-12 Steam Generatr,r and SRP Revision Reactor Cooli,nt Pump 5.3.4 Supports
- Complete BC - 50 changes necessary
- See pages (6) through (16) of itis Enclosure for remarks.
RA - Not appilcable I - Incomplete E - Evaluating actions required (2)
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-BSI/MPA WINEBER TITLE REF. DOEINEET APPLICMILITY
~ STATRS/DATE*
RDWWES**
A-17 Systems Interactions Letter: DeYoung to All I/03-31 A-17 Coolant Systems licenses - 09/72 NUREG-1174. NUREG-1229, NUREG/CR-4261,'NUREG/CR-4470, GL 89-18 NYD-89136 A-24 Qualifications of Class MUREG-0588. Rev. 1 All C/05-86 A-24 MPA B-60 IE Safety-Related SRP 3.11 Equipment 10 CFR 50.49 GL 82-09 GL 84-24 GL 85-15 SBN-1127 A-26/
Reactor Vessel Pressure DOR Letters to PWR C/11-85 A-26 r.PA B-04 Transient Protection Licensees 08/76 NUREG-0224 NUREG-0371 SRP 5.2 GL 88-11 NYN-88155 A-31 Residual Heat Removal MUREG-0606 All Ols After N/C A-31 Shutdown Requirements-RG 1.113. RG 1.139, 01/79 SRP 5.4.7 SBN-1105 A-36/
Control of Heavy Loads NUREG-0612 All C/06-04-86 A-36 C-10 Near Spent Fuel SRP 9.1.5 C-15 GL 81-07, GL 83-42, GL 85-11 Letter from DG Eisenhut dated 12/22/80
Complete RC - No cleanges necessary
- See pages (6) through (16) of the Encksure for remerts.
NA - Not appilcable I - Incomplete E - Evaluating actions required.
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- December; 11; 1989' j{,)
USI!MPA' NtNEBEk TITLE REF. DOCWEET APPLICABILITY STATES / BATE
- REMpaKS**
1 A-39 Determination of SRV NUREG-0802 BWR s/A.
Pool Dynam7c Loads NURF8.t,.0763, 0783, 0802, NUREG-0661 SRP 6.2.1.1.C A-40 Seismic Design SRP Revisions, MUREG/
All N/C A-40 Criteria CR-4776,'NUREG/CR-0054 NUREG/CR-3480, MUREG/
CR-1582, NUREG/CR-1161 N3EG-1233, NUREG-4776 NUREG/CR-3805 NUREG/CR-5347-j MUREG/CR-3509 A-42 Pipe Cracks in Bolling NUREG-0313 Rev. 1 BWR N/A MPA 8-05 Water Reactors MUREG-0313. Rev. 2 GL 81-03, GL 88-01 A-43 Containment Emergency NUREG-0510 All N/A A-43 Sump Performance NUREG-0869, Rev. 1 NOREG-0897, R.G.1.82 (Rev. 0), SRP 6.2.2 4
GL 85-22 (No Requirements)
A *4 Station Blackout RG 1.155 All C/04-17-89 A-44 NUREG-1032 MUREG-1109 10 CFR 50.63 NYM-89038
- Complete NC - No changes necessary
- See pages (6) tW (16) of this Enclosure for remerts.
RA - Bot applicable I - Incoglete E - Evaluating actions required (4)
,c Dr'. ester 11, 1969
't,j WSI/IFA ENBEER TITLE REF._D @ BT APPLICJBILITY STATW5JERTE*
M N W 5**
-A-45 Shutdown Decay Heat SECY 88-260 All I/03-31-91
.A-45 Removal Requirements NUREG-1289 NUREG/CR-5230 SECY 88-260 NTN-89136 (No requirements)
A-46 Seismic Qualification NUREG-1030 All N/A A-46 of Equipment in MUREG-1211/
Operating Plants GL 87-02. GL 87-03 A-47 Safety Impilcation NUREG-1217. NUREG-1218 All N/C A-47 of Control Systems GL 89-19 A-48 Hydrogen Control 10 CFR 50.44 All, except PWRs with N/A A-48 Measures and Effects SECY 89-122 large dry containments of Hydrogen Burns on Safety Equipment A-49 Pressurized Thermal RGs 1.154, 1.99 PWR C/07-08-88 A-49 Shock SECY 82-465 SECY 83-288 SECY 81-687 10 CFR 50.61/
NYN-88090
- Complete RC - se changes necessary
- See pages (6) threagh (16) of this Enclosure for remerts.
54 - met appilcable I - Incomplete E - Evaleeting actions required (5)
F December 11, 1989 l
USI A-1 Water Hammer This USI was resolved in March, 1984 by NUREG.0927. Revision 1.
No design modifications were imposed on operating plants or plants granted a i
construction permit prior to March, 1984.
A part of the resolution of this USI is the NRC revision of Standard Review Plan (SRP) Sections 3.9.3, 3.9.4, 5.4.7. 6.3, 9.2.1, 9.2.2, 10.3 and 10.4.7.
to provide specific water hammer-related review guidance for affectei systems and componente (RCS, RHRS, ECCS, Service Water, auxiliary cooling main steam, feedvater and condensate systems). Because these SRP revisions were not backfitted to apply to Seabrook, the SER and its supplements do not address water hammer for all the above systems. But, where water hammer has been specifically addressed in the PSAR, the SER has found the design acceptable.
FSAR Gection 10.4.7.3 discusses the design features of the steam generator and main feedwater piping that are designed to prevent water hammer.
Seabrook committed to perform emergency feedwater system flow stability tests to verify that unacceptable feedwater system water hammer will not occur.
Based on the commitment to perform these tests, SER Section 10.4.7 found that SRP 10.4.7 criteria had been met.
The emergency feedwater system flow stability tests were performed in October 1985. Water hammer was experienced in the steam supply lines to the turbine driven emergency feedwater pump.
The design of the steam supply lines to the turbine driven emergency feedwater pump was modified.
Subsequent testing confirmed that no e
further water hammer problems existed.
Although water hammer was not specifically mentioned in FSAR Section 6.3, SER Section 6.3.1 discusses the acceptability of ECC6 systems relative to water hammer based upon the commitment to maintain lines filled.
Two licensee actions are assumed by NRC to be part of the resolution of this USI. First, the continuing review of operating experience in accordance with NUREG 0737. Item I.C.5.
Second, the administration of training programs in accordance with NUREG 0737. Item I.A.2.3.
With regard to operating experience review. FSAR Sections 1.9 and 13.5.1 indicate that Seabrook will prepare procedures for feedback of operating experience to plant staff th'at meet the requirements of NUREG 0737. Item I.C.5.
An approved NHY procedure requires an operating experience review program addressing NUREG 0737. Item I.C.5 requirements.
SSER4 lists implementation of NUREG 0737 Item I.C.5 as Confirmatory Issue #44.
Verification of Completion of Confirmatory Issue #44 has been delegated to the NRC Region I Inspection Staff in SSERS.
With regard to the administration of training programs, FSAR Sections 1.9 and 13.2.1 address the Seabrook commitment to NUREG 0737. Item I.A.2.3 in general terms.
SER, Section 13.2.1.3 concludes that this NUREG 0737 item has been complied with.
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USI A - 1 (continued)
I Therefore, implementation of this USI is complete (C) as of Jur.e 1989, the date of completion of subsequent testing of the re-designed steam supply lines to the turbine driven emergency feedwater pump.
USI A - 2 Asymmetric Blowdown Loads on Reactor Primary Coolant Systems This USI was resolved in January, 1981 by NUREG 0609 which required designed protection against the asymmetric loads resulting from postulated, i
rapid, double-ended breaks in RCS piping. FSAR Section 3.9(N)2.5 describes the methodology for developing dynamic loads from postulated pipe breaks.
SER, Section 3.9.2.4 indicates that this analysis methodology meets the guidelines of NUREG-0609.
In Section 3.6.2, the SER indicates that the requirements of 10CFR50, Appendix A General Design Criterion GDC-4 have r
been met.
Generic Letter (GL) 84-04 permitted licensees to request an exemption to the requirements of GDC-4 with respect to asymmetric loads.
Licensee requirements to support the exemption should include:
1.
The appropriate ' leak before break' unalysis.
2.
Leakage detection systems as described by GL 84-04.
The GDC-4 exemption request for Seabrook Station was submitted by PSNH letters SBN-703 (08/09/84) and SBN-750 (02/01/85).
In SSERS, the GDC-4 exemption request was approved until completion of the second refueling outage, pending the outcome of rulemaking. GDC-4 was revised by rulemaking on October 27, 1987.
FSAR Section 5.2.5 describes leakage detection systems that meet the guidelines of Regulatory Guide 1.45 (as discussed in GL 84-04).
SER Section 5.2.5 finds the Seabrook Leakage Detection Systems acceptable.
Therefore, implementation of this USI is complete (C) as of February, 1985, the date of the second letter (SBN-756) requesting the GDC-4 exemption.
USI A-3 Westinchouse Steam Generator Tube Integrity This USI was resolved in September 1988 by NUREG - 0844. No generic requirements were imposed as a part of the resolution of this USI.
However, staff recommended actions were identified. These staff recommended actions were cost-beneficial enhancements that licensees could voluntarily implement to further reduce public risk and provide added assurance that risk will continue to be small.
NRC Generic Letter (GL) 85-02 issued the staff recommended actions identified in NUREG-0844. PSNH Letter SBN-815 dated June 13, 1985 was (7)
5 December 11, 1989 USI A - 3 (continued) submitted in response to GL 85-02.
New Hampshire Yankee (NEY) is currently verifying the completion status of commitments made in SBN-815.
Therefore the implementation status of this USI is listed as incomplete (I) pending verification of the completion of status of commitments made in SBN-815.
USI A - 9 Anticinated Transients Vithout Scram (ATWS)
This USI was resolved by 10CFR50.62, which required each PWR to install equipment, diverse from reactor trip equipment, to trip the turbine and initiate emergency feedwater under conditions indicative of ATWS.
In letters, NYN-88161 (12/15/88) and NYN-89027 (03/15/89), New Hampshire Yankee (NHY) described the ATWS Mitigation System for Seabrook Station.
NRC provided a Safety Evaluation Report on June 15, 1989 concluding that the ATVS design proposed by NHY was in compliance with 10CFR50.62 requirements.
In letter NYN-89084 (07/06/89), NHY confirmed that installation of the ATWS Mitigation System would be completed on or before August 31, 1990.
Thus the implementation status of this USI is incomplete (I) pending installation of the ATWS Mitigation System.
USI A - 11 Resctor Vessel Materials Tounhness This USI was resolved in October, 1982 by NUREG-0744 which stated that: "In accordance with the requirements of 10CFR50, Appendix G, all licenses should take the following course of action.
The USE* at the plant-specific end of life (EOL) should be established in accordance with 10CFR50 and the ASME Code. If the EOL USE2 50 ft-lb, the RPV is acceptable (other factors, detailed in 10CFR and in the Code, remain in force).*
FSAR Section 5.3.1.5 indientes that the Seabrook Station reactor pressure vessel material meets the initial fracture toughness requirements of i
10CFR50, Appendix G and will be evaluated to ensure that Appendix G requirements are met for the life of the plant.
SER Sections 5.3.1 and 5.3.3 indicate that the Seabrook reactor pressure vessel meets 10CFR50, Appendix G requirements and thus additional analysis per NUREG-0744 is not required.
Therefore, this USI is applicable to Seabrook but no changes are necessary (N/C).
USE is the reactor pressure vessel material sample Charpy V-notch test upper shelf energy level.
(8)
1 December 11, 1989 USI A - 12 Potontial of Low Fracture Touthness and Lamellar Tearina
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in PWR Steam Generator and Reactor Coolant Pumo Suenorts This USI was resolved in October 1983 by NUREG 0577, Revision 1.
NUREG 0577 Revision 1 states that 'The fracture toughness of materials in existing support structures is adequate for those...that meet the requirements of Paragraph NF 2311, ASME Code,*
NUREG 0577, Revision 1 states further that modifications to existing support structures could not be justified and that requirements to certify the acceptability of material or designs should not be imposed.
l The implementing action to resolve this USI was the revision of the Standard Review Plan (SRP)
Section 5.5.4.
The revised SRP review guidance would be applicable only to new construction.
Since Seabrook Station was l
1ssued a Construction Permit on 1976, this USI is not applicable (N/A) to Seabrook.
Seabrook design, nonetheless, meets the criteria of NUREG 0577. Revision 1.
FSAR Section 5.4.14 states that component supports are designed in accordance with the ASME Code Section III Subsection NF, except as noted i
in Subsection 3.8.3.
SER Appendix C, recognizes the design of the component supports ir. accordance with the ASME Code,Section III Subsection NF, which ensures adequate fracture toughness.
USI A - 17 Systems Interactions in Nuclear Power PlsD.$.J!.
This USI was resolved in September, 1989 by Generic Letter (GL) 89-18.
As part of the resolution of this USI, NRC anticipated that licensees would take the following actions:
1)
Consider insights provided in NUREG 1174 regarding plant vulnerabilities to flooding and water intrusion from internal sources in performing the Individual Plant Examinations for Severe Accident Vulnerabilities (IPE).
IPEs were requested by Generic Letter 88-20.
2)
Continue to review operating experience informa-tion in accordance with NUREG 0737, Item I.C.5.
Action (1) will be completed as part of the IPE scheduled to be completed by March 31, 1991 (Ref: NHY letter, NYN-89136).
Action (2) is ongoing in accordance with the NHY Operating Experience Review Program required by an approved NHY procedure.
Therefore, the implementation status of this USI is incomplete (I) pending submittal of the IPE report on or before 03/31/91.
(9)
December 11, 1989 USI A - 24 qualification of Class IE Eauinment l
This USI was resolved in July, 1981 by KUREG-0588 Revision 1.
Generic Letter 89-21. Enclosure 2, Page 7 states the following:
'In summary, the resolution of A-24 is embodied in 10CFR50.49. 'A measure of whether each licensee has implemented the resolution of A-24 may, therefore, be found in the determination of compliance with 10CFR50.49.'
FSAR, Section 3.11 discusses the environmental design and qualification of mechanical and electrical equipment.
i SER Section 3.11 restates the requirements for environmental qualification of Class 1E equipment.
The subject is addressed substantially by the NRC in SSER$. SSER5 documents the review of eleven NHY submittals, an audit conducted in February 25-26, 1986, and concludes:
'On the basis of the results of its review and subject to confirmation that all audit deficiencies have been corrected, the staff concludes that the applicant has demonstrated compli-ance with environmental qualification as outlined in 10CFR50.49, the relevant parts of GDC 1 and 4, and Sections III, XI, and XVII of Appendix B to 10CFR50, and with criteria as specified in NUREG-0588.'
The audit deficiencies referred to in the above quotation were documented in a Meeting Summary published by the NRC on April 11, 1986.
NHY documented closure of all theso deficiencies in an NRC EQ Completion Letter dated June 20, 1986.
In SSER6, the NRC cited this correspondence and stated that they considered SER Confirmatory Item 47 (Environmental Qualification of Equipment) closed.
Therefore, implementation of this USI is considered complete (C) as of June 20, 1986.
USI A - 26 Reactor Vessel Pressure Transient Protection This USI was resolved in 1970 by NUREG 0224 and Standard Review Plan (SRP),
Section 5.2.
SRP Section 5.2 contains Branch Technical Position RSB 5-2 which specifies the design requirements for a Low-Temperature overpressure Protection System.
FSAR Section 5.2.2.11 describes the Seabrook Station Low Temperature Overpressure Protection (LTOP) System. SER Section 5.2.2.2 finds the Seabrook Station LTOP System acceptable.
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o December 11, 1980 USI A - 26 (continued)
Therefore, implementation of this USI is complete (C) as of November 26, 1985, the date of submittal of FSAR Amendment $56.
(Ref: PSNH Letter SBN-900).
Generic Letter (GL) 88-11 is listed by NRC as a reference document to this USI. NHY letter NYN-88155, (11/30/88) responded to GL 88-11 and committed to a future Technical Specification change. This planned Technical Specification change will affect the Pressure-Temperature Operating Curves and, as a result, may affect the LTOP System setpoint. At the time of submittal of the planned Technical Specification change, the impact of the change on the LTOP System setpoint will be considered.
Therefore, this planned Technical Specification change does not affect the completion status of this USI.
USI A - 31 Residual Heat Removal Shutdown Reauirements This USI was resolved in May, 1978 by Standard Review Plan (SRP), Section 5.4.7.
Seabrook Station is required to meet BTP RSB 5-1 as a Class 2 plant for implementation (partial implementation of SRP, Section 5.4.7 is required).
FSAR Section 5.4.7 describes the RHR System. The SER, SSER5 and SSER6 discuss the review of the RHR System and find it acceptable.
Therefore, this USI is applicable to Seabrook but no changes are necessary (N/C).
USI A - 36 Control of Heavy Loads Phases I and II This USI was resolved in July, 1980 by NUREG-0612 and Standard Review Plan (SRP). Section 9.1.5.
PSNH has provided information in the letters listed below regarding conformance of Seabrook Station to the guidelines of NUREG-0612.
This information demonstrates that the intent of the guidelines of NUREG-0612 Section 5.1.1 (the Phase I response) have been met at Seabrook Station.
NRC review of the implementation of the NUREG-0612 Phase I guidelines is documented in SER and SSER5. The conclusion of SSERS is that the implementation of the response to NUREG-0612, Phase I guidelines is acceptable. Generic Letter (GL) 85-11 indicated that, as a result of the Phase I responses, no further responses associated with Phase II were required.
Therefore, implementation of this USI is considered complete (C) as of June 4 1986, the date of the final response letter to NRC on the subject.
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USI A - 36 (continued) i List of written submittals regarding NUREG-0612:
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1 PSNH Letter No.
Date I
SBN-331 September 24, 1982 SBN-418 January 6, 1983 SBN-521 June 20, 1983
.SBN-594 December 15, 1983 SBN-872 September 20, 1985 SBN-887 October 31, 1985 SBN-977 March 31, 1986 SBN-1088 June 4, 1986 USI A - 40 Seismic Design Criteria This USI was resolved in June 1989 by NUREG-5347.
In resolving this USI, NRC will revise Standard Review Plan (SRP) Sections 2.5.2, 3.7.1, 3.7.2 and 3.7.3.
NUREG 1235 states that only plants with Construction Permits docketed after the date of issuance of the revised SFP Secticns will be subject to the new review criteria. NUREG-1233 further states that plants r
subjected to licensing review since 1984 have been confirmed by the NRC Staff to have appropriate above-ground tank design.
i The subject of tank design criteria was addressed at Seabrook during the 1983 Integrated Design Inspection when the Refueling Water Storage Tank (RWST) design was reviewed. PSNH addressed this subject in letters SBN-678 (06/29/84) and SSN-913 (12/26/85). Seismic Category 1 tanks were re.
analyzed in accordance with NUREG/CR-1161 ' Recommended Revisions to Seismic Design Criteria.' The re-analysis results for each tank were then compared to the tank design calculations performed by the tank vendors and, in each j
case found acceptable.
Although the other SRP changes concerning (a) clarification of development of site specific spectra (b) justification for use of single synthetic time
- history by power spectral density function and (c) location and reductions of input ground motion for soil structure interaction are not being applied to plants with current licenses or applications, SER Appendix C indicated that "The Staff does not expect the results of Task A-40 to affect these l
conclusions because the techniques under consideration are essentially l
those utilized in the review of this facility."
l Therefore, the USI is applicable to Seabrook, but no changes are necessary I
(N/C).
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.s December 11, 1989 USI A - 43 Containment Emergency Sumn Performance This USI was resolved in October, 1985 by NUREG-0869. Revision 1 and NUREG.
0897, Revision 1.
NUREG-0869. Revision 1 concluded that no backfit requirements were require 1 for pressurized water reactor PWR dry containments (5.2).
NRC revised Regulatory Guide 1.82 and Standard Review Plan (SRP) Section 6.2.2, and issued Generic Letter (GL) 85-22.
GL-85-22 states that the revised guidance shall not apply to plants operating or under construction-as of the end of 1985.
Therefore, this USI is not applicable to Seabrook (N/A).
Although the revised guidance developed to resolve this USI does not apply to Seabrook, considerable action has been taken to address the issues presented by this USI. A scale model test program for the Seabrook Containment Emergency Sump was conducted to study intake head losses and
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vortex control. During the tests utilizing this model, the observed vortexing was limited to an acceptable level. It was also observed that up to fifty percent blockage of the sump screen area did not degrade sump performance, t
FSAR Section 6.2.2.2 discusses the design features of the containment sump.
The types of insulation utilized in containment and the potential for this insulation to cause sump blockage is discussed. The insulation used is designed to remain in place during a seismic event and not be adversely affected by postulated accident containment environments. Any insulation dislodged by a LOCA or high energy line jets would settle out on the containment floor and not be transported to the sump.
SER Section 6.3.1 finds insulation design acceptable.
PSNH addressed the effects of high energy line breaks in the vicinity of the containment sump.
PSNH letter SBN-370 (11/15/82) provided information indicating that high energy lines in the vicinity of the sump could be isolated in the event of a pipe break. SER Section 6.3.1 finds this response acceptable.
SER Appendix C discusses USI A-43 relative to Seabrook and concludes that the containment sump design is adequate.
USI A - 44 Station Blackout This issue was resolved by the NRC in June 1988 by 10CFR50.63 and Regulatory Guide 1.155.
NHY provided the required response to 10CFR50.63 on April 17, 1989 in letter NYN-89038.
The NHY response provided the proposed station blackout duration and described the results of the coping assessment. NHY determined that no (13)
Ti December 11, 1989
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USI A - 44 (continued) modifications to existing equipment were required in order to comply with 10CFR50.63.
The necessary procedure revisions were identified in the response and have been completed.
Therefore implementation of this USI 10 considered complete (C) as of April 17, 1989.
USI A - 45 Shutdown Decay Heat Removal Recuirements This USI has been resolved by including it within the Individual Plant Examination for Severe Accident Vulnerabilities (IPE) Program.
In letter NYN-89136, NHY indicated that the Seabrook Station IPE would be completed by March 31, 1991. Therefore, the implementation status of this USI is incomplete (I) pending submittal of the IPE report.
In_1983, the Seabrook Station Probabilistic Safety Assessment (SSPSA) was completed.
In 1988, the Seabrook Station Probabilistic Safety Study for Shutdown Modes 4, 5 and 6 was completed.
The 1988 Shutdown Modes Study effectively increased the scope of the SSPSA to include evaluation of plant operation in the shutdown modes.
The IPE will be based upon these two existing studies.
USI A - 46 Seismic Oualification of Eauipment in Operatina Plant This USI was resolved in February, 1987 by Generic Letter (GL) 87-02.
The purpose of this USI was to investigate the adequacy of seismic qualification of mechanical and electrical equipment installed in older nuclear power plants. The specific plants subject to USI A-46 are listed in NUREG-1211.
The list of applicable plants does not include Seabrook l
Station.
GL 87 02 described the acceptable seismic verification methodology for the applicable plants. Generic Letter (GL) 87-03 was addressed to Seabrook in lieu of GL 87-02.
GL 87-03 stated:
'We have documented evidence in staff SERs that your plant either has been, or is required to be, reviewed to current licensing requirements for Seismic Qualifications of Equipment (i.e., SRP-3.10 IEEE-344/75 and Regulatory Guide 1.100) and therefore you are not required to respond to this letter or to perform the plant reviews described in the enclosures."
SER Apper. dix C states:
'This USI pertains specifically to operating plants, particularly those older plants where seismic qualification was accomplished before the development of current licensing requirements.'
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9 4 4 December 11, 1989 USI A - 46 (continued)
I FSAR Section 3.9(B).2.2 describes the seismic design criteria for Cate-gory 1, safety-related mechanical equipment.
FSAR Section 3.10(B) de-scribes the seismic qualification of Category 1 electrical equipment and instrumentation.
Therefore, based upon the above, this USI is not applicable to Seabrook (N/A).
USI A - 47 Safety Imolications of Control Systems in LWR P2 clear Power Plants This USI was resolved by Generic Letter.(GL) 89-19.
GL 89-19 recommends automatic steam generator overfill protection, Technical Specifications and procedures to periodically test and verify the operability of this feature.
New Hampshire Yankee has conducted a review of Seabrook Station design.
Technical Specifications and surveillance procedures to verify that xisting steam generator overfill protection meets the recommendations of GL 89-19.
New Hampshire Yankee has concluded that Seabrook Station has the recommended automatic protection. Technical Specifications and surveillance procedures.
Therefore this USI is considered applicable but no changes are required (N/C).
A response to Generic Letter 89-19 is required by March 19, 1990.
USI A - 48 Hydronen Control Measures and Effects of Hydronen Burns on Safety Eautoment Generic Letter 89-21 indicates that this USI is not applicable to pressurized water reactors with large. dry containments.
The Seabrook Station containment'is a large, dry containment. Therefore, this USI de not applicable to Seabrook (N/A).
SER Appendix C states that the applicability of USI A-48 is restricted to Bolling Water Reactor (BWR) containments and containments of ice condenser design.
USI A - 49 i
Pressurized Thermal Shock l
This USI has been resolved by 10 CFR 50.61 which established a screening criteria for specific reactor pressure vessel sections and welds.
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F December 11, 1989 USI A - 49 (continued)
On July 8, 1988 NHY submitted letter NYN-88090 which contained information indicating that the Seabrook reactor pressure vessel would meet the PTS screening criteria for the life of the plant.
Therefore, implementation of this USI is complete (C) as of 07/08/88.
Generic Letter (GL) 88-11 is listed by NRC as a reference document to this USI.
In letter NYN-88155, dated 11/50/88 NHY, responded to GL 88-11 and committed to a future Technical Specification change. This future Technical Specification change does not impact the completed status of USI A-49 for the following reason.
In letter NYN-88155, NHY committed to Revision 2 of Regulatory Guide 1.99.
Revision 2 of Regulatory Guide 1.99 revises pressure-temperature curve calculation methodology but does not affect the 10CFR50.61 screening criteria.
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