ML17334B337
| ML17334B337 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 12/07/1989 |
| From: | Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TASK-***, TASK-OR AEP:NRC:1108A, GL-89-21, NUDOCS 8912140402 | |
| Download: ML17334B337 (99) | |
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Indiana Michigan Power Company P.O. Box 16631 Cofumbus, OH 43216 INDIANA NICHIGAN POWER AEP:NRC:1108A Donald C.
Cook Nuclear Plant Units 1 and 2
Docket Nos.
50-315 and 50-316 License Nos.
DPR-58 and DPR-74 NRC GENERIC LETTER 89-21:
ADDITIONAL INFORMATION U.
S, Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.
C.
20555 Attn:
T. E. Murley December 7,
1989
Dear Dr. Murley:
This letter and its attachment provide additional information and revised pages of attachment to our previous letter AEP:NRC:1108 dated November 29, 1989.
The revisions are being made as a result of our conversation with the NRR Project Manager.
This document has been prepared following Corporate procedures that incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.
Sincerely, M. P. Alexich Vice President MPA/eh Attachment cc:
D. H. Williams, Jr, A, A. Blind - Bridgman RE CD Callen G
~ Charnoff NFEM Section Chief A. B. Davis - Region III NRC Resident Inspector
- Bridgman
ATTACHMENT, TO AEP 08 UNRESOLVED SAFETY 1SSUES FOR WHICH A FINAL TECHNICAL RESOLUTION HAS BEEN ACHIEVED US I/NPA RIRRIER A-l TITLE Mater Haplner REF.
DOCUHENT SECY 84-119 NURECR-09?7, Rev.
1 NUREG-0993; Rev.
I NUREG-0737 Item I.A.?.3 SRP revisions APPL ICABILITY A11 STATUS DATE%
NA AEP:NRC:0398 and 0678 Series of letters RERRRKS Based on enclosure 2 to GL 89-21 A-2/
Asymmetric 8 Iowdown KPA 0-10 Loads on Reactor Primary Coo1ant Systems NUREG-0609 GL 84-04.
GOC-4 PMR CAEP:NRC:0137D Dated: Sept.
10, 1984 NRC SER dated Nov. 22, 1985 Amendment No. 76 to Operating License No. DPR-74.
A-3 A-I A-5 EI-6 Mestinghouse Steam Generator Tube Integrity CE Steam Generator Tube Integrity BSM Steam Generator Tube Integrity Nark I Containment Short-Term Program NUREG-0844 SECY 86-9?
SECY AR-?72 GL 85-02 (No requirements)
NUREG-0844, SECY 86-97 SECY 88-272 GL 85-02 (No requirements)
NURFG-0844.
SECY 06-97 SFCY 88-272
'L 85-02 (No Requirements)
'I4-PMR CE-PMR ngw-pwfr Hark I-BMR C
AEP:NRC:0936 dt. June 2l, 1985
-NA-
-NA-Cook Unit 2 steam generators were replaced; for Unit 1 additional information is being provided to NRC under IE Bulletin 88-02 None None None i C - CON'LETE NC - NO CHAN6ES NECESSARY NA - IIT APPLICABLE I - INCONPLETE E - FVALUATIN6 ACTIONS RE(UIRED
ATTACHMENT TO ART:
08-US l/MPA KBKBER A-7/
O-O)
T JTLE Nark I Long-Term Program
?
REF.
DOCUHFNT NUREG-0661 NUREG-0661.uppl.
I GL 79-57 APPL fCABILITY Hark
)-81IR STATUS DATE*
-NA-REKRRRB None A-8 Mark fl Containment Pool Dynamic Loads NUREG-0808 NUREG-048,", Supp).
I/?
NUREG-0802 SPP 6.7..). IC GOC 1&
Marl I ) -81IR
-NA-None A-9 Anticipated Transients Without Scram A-10/
BIBJR Feedwater Nozz)e MPA B-25 Cracking NUREG-0460, Vof.
4 10 CFR 50.67" NUREG-0619 Letter from OG EisEnhut dated 11/13/80 GL 81-) I A) I C
AEP:NRC:0838 Series of letters
-NA-None A-11 A-)7 A-17 A-24/
I~PA B-I Reactor Vessel Materia)
Touohnes s Fracture Toughness of Steam Generator and Reactor Coo)ant Pump Supports Systems Interact ions n ;~ Eication of C)ass
. ty-Re)ated cqu>pment
. NUREG-0744, Rev.
1 10 CFR 50.60/
8?-26 NUREG-0571, Rev.
l SRP Revision 5.3.4 Ltr:
Oeyoung tu
)icensees
- 9/7?
IIUREG-) 174, NUREG-
- 1224, NUREfi/CR-392?,
NUREG/CR-4?61, NUPEG/
CR-4470, GL 89-18 (No requirements)
NIJREG-0588, Rev.
I SRP 3.11 10 CFR 50.49 GL 8--04, GL 84-24 GL 85-)I)
A))
PXP A>)
A) I
-NC-See note in cover letter, Ztem l Our letters-AEP:NRC:0356 Series AEP:NRC:0578 Series AEP:NRC:0775 Series ATTACHMENT TO AEB-08 USI/NPA IINIBER A-26/
NPA 8-04 A-3!
TITLE Reactor Vessel Pressure Transient Protection Pesidual Heat Removal Shutdown Requirements REF.
OOCINENT OOP. Letters to Licensees 8/76 NUREG-0224 NUREG-0371 SRP 5.2 GL 88-11 NUREG-0606 RG 1.113, RG 1.139 SRP 5.4.7 APPL ICABILITY PMR Al 1 OLs After 01/79.
C See cover letter, Item 2 AEP:NRC:0894K T/S specification changes dt.
Dec.
5, 1988 are in progress AEP:NRC:0894L dt.
Oct.
25, 1989 NRR Inspection Report 50-315; 316/86-34 dt. 11/4/86
-NA-A-36/
C-IO, C-15 A-39 Control of Heavy Loads Near Spent Fue>
Determination of SRV Pool Dynamic Loads and Pressure Transients NUREG-0612 SRP 9.1.5 GL BI-IP7, GL 83-42, GL 85-I 1 Letter from OG Eisenhut dated 12/22/80 NUREG-0802 NUREGs-0763,0783,0802 NUREG-0661 SPP 6.2.1.1.C Al I BMR C
AEP:NRC:0514 Series of letters NRC SER (Phase I) dt. Sept.
20, 1983
-NA-See cover letter, Item 3 None A-40 Seismic Design Criteria SRP Revisions, NURLG/
All CR-4776, NUREG/CR-0054, NUREG/CR-3480, NUREG/
CR-1582, NUREG/CR-1161, NUREG-1233, NUREG-4776 NUREG/CR-3805 NUREG/CR-5347 NUREG/CR-3509
-NC-See cover letter, A-42/
Pipe Cracks in Boiling 11PA 8-05 Mater Reactors NUREG-0313, Rev.
1 NUREG-0313, Rev.
-NA-None
ATTACHMENT TO AEP:
108'4-USI/jIPA NlNBER TITLE REF.
00CUMENT APPLICABILITY STATUS DATE*
RBNRKS A-43 A-44 Containment Emergency Sump Performance Station Blackout NUREG-0510, KUREG-0869.
Rev.
I NUREG-0897, R.G. I.BP (Rev. 0),
SRP 6.2.2 GL 85-22 No Requirements RG I.I55 NIiREG-1032 KUREG-1109 10 CFR 50.63 I, our letter AEP:NRC:0537D dt. April 14, 1989 No NRC submittal is required Awaiting NRC review A-45 A-46 A-47 Shutdown Oecay Heat Removal Requirements Seismic Oualification of Equipment in Operating Plants Safety Imp>icatinn of Control Systems SECY RR-260 NUREG-1289 NUREG/CR-5230 SECY 88-260 (No requirements)
NIIREG-103r.
NUREG-1211/
GL 87-02, Gl 87-03 NUREG-I II, NURrr 12IR rL eg-ig Al I C, AEP:NRC:1082 dt. Oct.
24, 1989 AEP:NRC:1040 I
This has been incorporated into IPE program under GL 88-20 See cover letter
> Item 5 Under review and evaluation by AEPSC A-48 Hydrogen Contro I Neasures and Effects of Hydrogen Burns on Safetv Equipment 10 CFP 50.44 SECY 89-122 All, except PIIRs with large dry contaieaents I
Remaining open items will be AEP:NRC:0476 and addressed in the IPE program AEP:NRC:0500 Series by AEPSC A-49 Pressurized Thermal Shock RGs 1.154, I.gg SFrv 82 465 SECY 83-288 SECY 81-687 10 CFR 50.61/
GL 88-11 PMR C~
AEP NRC:0561A NRC SER dated March 27, 1987 dt. January 22, 1986
ENCLOSURE 2 USI STATUS
SUMMARY
PLANT D.C. Cook Units 1 and 2
DOCKET NO(S).
50-315 AND 50-316 PROJECT MANAGER J.
G. Giitter TECHNICAL CONTACT Jai-Ra,ian USI NO.
A-2 -.
TITLE As mmetric Blowdown Loads in RCS MPA NO.
D-10 TAC NOS.
08477.and 08478 ISSUES
SUMMARY
This USI was resolved in January 1981 with the publication of NUREG-0609, "Asymmetric Blowdown Loads on PHR Primary Systems."
In October 1975, the NRC notified each operating PWR licensee of a potential safety problem concerning the fact that asymmetric LOCA loads had not been considered in the design of any PWR piping system.
In June 1976 the NRC informed each PWR licensee that it was required to reassess the reactor vessel support design of its facility.
The staff expanded the scope of the problem in January 1978 with a request for additional information to all PHR licensees.
NUREG-0609 provided guidance for these analyses.
For operating PWRs, Multi-Plant Action (MPA) Item D-10 was established by NRC's Division of Licensing for implementation purposes.
During the course of the work on USI A-2, it was demonstrated that there were only a very limited number of break locations which could give rise to significant loads.
Subsequently, after substantial new technical work, it was demonstrated that pipes would leak before break and that new fracture mechanics techniques for the analyzing of piping failures assured adequate protection against failures in primary system piping in PHRs (Generic Letter 84-04).
This was reflected in a revision of General Design Criteria (GDC)-4 (Appendix A to 10 CFR Part 50) published in the Federal Re ister in final form on April ll, 1986, and in a subsequent revision to GDC-4 pu ss ed in the Federal Re ister on July 23, 1986.
In addition, it has also been
~sat>s actor> y emonstrated in the course of the A-2 effort that there is a
very low likelihood of simultaneous pipe loading with both LOCA and safety shutdown earthquake (SSE) loads.
Therefore, the last revision of GDC-4 represented the final technical action of NRC regarding the issue of asymmetric blowdown loads issue in PWRs primary coolant main loop piping.
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIFIC):
On February 1, 1984, the staff issued Generic Letter 84-04 a safety evaluation of the Westinghouse topical reports dealing with elimination of postulated pipe breaks in PWR primary loops.
The safety evaluation concluded that an acceptable basis had been provided so that asymmetric blowdown loads
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1 resulting form double ended pipe breaks in main coolant loop piping need not be considered as a design basis for the Mestinghouse Owner's Group plants (including D. C. Cook) provided that two conditions are met.
The first condition, which did not apply to Cook, involved verification of bending moments at two other nuclear power plants.
The second condition was that leakage detection systems exist to detect postulated flaws utilizing guidance from Regulatory Guide 1.45 (with the exception of seismic equipment qualification for the airborne particulate radiation monitor).
At least one leakage detection system sensitive enough to detect a
1 gpm leak within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> was required to be operable.
The licensee responded in letter dated September 10, 1984, indicating that the leak detection systems at Cook are consistent with the requirements of Generic Letter 84-04.
In this same letter the licensee proposed to remove license condition C.3(a) to Operating License No. DPR-74 for Unit No. 2, which requires that an analytical evaluation be made of the effects of certain postulated break, loads on the reactor coolant system and internals.
This license condition was removed by Amendment No.
76 dated November 22, 1985.
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REFERENCES:
1.
RE UIREMENT DOCUMENTS:
Cook Units 1 and 2
A-2 NUDOCS NO.
TITLE Generic Letter "Evaluation of Primary Systems for Asymmetric LOCA Loads" Task Action Plan A-2, "Asymmetric Blowdown Loads on Reactor Primary Coolant System,"
NUREG-0371 Task Action Plans for Generic Activities DATE 01/20/78 11/78 "Asymmetric Blowdown Loads on PWR Primary Systems,"
NUREG-0609 US NRC NRR 01/81 GDC-4, "Environmental and Dynamic Effects Design Basis" GL 84-04, "Safety Evaluation of Westinghouse Topical Reports Dealing With Elimination of Postulated Pipe Breaks in PWR Primary Main Loops."
02/01/84 2.
If1PLEMENTATION DOCUMENTS:
TITLE NUDOCS NO.
DATE (1)
LTR from Alexich (AEP) to Denton 8409130354 9/10/84 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE
C I 1
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PLANT D. C. Cook-Units.1-and 2
PROJECT MANAGER J. G.-Giitter..
DOCKET NO(S).
50-315 and 50-316 TECHNICAL CONTACT J.
Mauck-USI NO.
A TITLE ATWS-er 10 CFR 50.62 MPA NO.
A-20 TAC NOS.
59082 and 59083 ISSUES
SUMMARY
This USI was resolved in June 1984 with the publication of a final rule (10 CFR 50.62) to require improvements in plants to reduce the likelihood of failure of the reactor protection system (RPS) to shut down the reactor following anticipated transients and to mitigate the consequences of an anticipated transient without scram (ATWS) event.
The rule includes the following design-related requirements:
50.62(C)(1),
diverse and independent auxiliary feedwater initiation and turbine trip for all PWRs; 50.62(C)(2), diverse scram systems for CE and BKW reactors; 50.62(C)(3) alternate rod injection (ARI) for BWRs. 50.62(C)(4);
standby liquid control system (SLCS) for BIIRs; and 50.52(C)(5),
automatic trip of recirculation pumps under conditions indicative of an ATWS for BWRs.
Information requirements and an implementation schedule are also specified.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC:
In response to paragraph 50.62 (C)(1), the Westinghouse Owners Group (WOG) developed a set of conceptual ATWS mitigating system actuation circuitry (AMSAC) designs for Westinghouse PWRs.
These designs were described in Westinghouse Topical Report, WCAP-10858, "AMSAC Generic Design Package".
The staff reviewed WCAP-10858 and issued a safety evaluation on July 7, 1986 indicating that the generic designs presented in the WCAP adequately meet the requirements of 10 CFR 50.62.
A revision to the WCAP involving a change in the AMSAC permissive signal was also found to be acceptable by the staff.
In a letter dated November 7, 1986 the licensee transmitted preliminary information on the detailed design of the AMSAC proposed for installation at Cook.
The licensee provided additional information related to the AMSAC design in letters dated June 25, 1987, October 28, 1987, December 18, 1987, March 31, 1988, and May 2, 1988.
The staff completed a safety evaluation of the Cook AMSAC design on July 1, 1988.
The safety evaluation concluded that the AMSAC design proposed for D.
C.
Cook is acceptable provided that electrical isolation devices are successfully qualified.
The safety evaluation was transmitted to the licensee in a letter dated April 14, 1989.
AMSAC was implemented for Unit 1 during the 1989 refueling outage which ended on July 8, 1989.
AMSAC was implemented for Unit 2 during the steam generator replacement outage which ended in February 1989.
NRC Inspection Report 89-032 concluded that the installation and testing of AMSAC at D. C.
Cook is acceptable.
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REFERENCES 1.
RE UIREMENT DOCUMENTS:
Cook Units 1 and 2
A-9 TITLE NUDOCS HO.
DATE NUREG-0460, and Supplements, "Anticipated Transients Without Scram for Light Water Reactors" Federal Register Notice 49 FR 26045 (10 CFR 50.62) 03/80 06/26/84 2.
IMPLEMENTATION DOCUMENTS:
TITLE Stang (NRC) to Alexich NUDOCS NO.
8904210297 DATE 4/14/89 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE Cook Inspection Report 0 89-032 TBD 11/13-14/89
PLANT D.C. Cook, Units-1 and-2 DOCKET NO(S).
50-315 and-50-316 PROJECT MANAGER J.
G. Giitter TECHNICAL CONTACT B. Elliott USI NO.
A-Il TITLE R
"-V 1~f1 i
1 T
h MPA NO.
A-07 ISSUES
SUMMARY
TAC NOS.
This USI was resolved in October 1982 with the publication of NUREG-0744, "Pressure Vessel Material Fracture Toughness.".
NUREG-0744 was issued by Generic Letter 82-26 and provided only a methodology to satisfy the require-ments of 10 CFR Part 50, Appendix G.
No licensee response to Generic Letter 82-26 was required.
Because of the remote possibility that nuclear reactor pressure vessels designed to the ASME Boiler and Pressure Vessel Code would fail, the design of nuclear facilities does not provide protection against reactor vessel failure.
Prevention of reactor vessel failure depends primarily on maintaining the reactor vessel material fracture toughness at levels that will resist brittle fracture during plant operation.
At service times and operating conditions typical of current operating plants, reactor vessel fracture toughness properties provide adequate margins of safety against vessel fai lure; however, as plants accumulate more and more service time, neutron irradiation reduces the material fracture toughness and initial safety margins.
Appendix G to 10 CFR Part 50 requires that the Charpy upper shelf energy throughout the life of the vessel be no less than 50 ft-lb unless it is demonstrated that lower values will provide margins of safety against fai lure equivalent to those provided by Appendix G of the ASME code.
USI A-11 was initiated to address the staff's concern that some vessels were projected to have beltline materials with Charpy upper shelf energy less than 50 ft-lb.
NUREG-0744 provides a method for evaluating reactor vessel materials when their Charpy upper shelf energy is predicted to fall below 50 ft-lb.
Plants wi 11 use the prescribed method when analysis of irradiation damage predicts that the charpy upper shelf energy is below 50 ft-lb.
IMPLEMENTATION=AND STATUS
SUMMARY
=-(PLANT SPECIFIC):
In their response to Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its impact on Plant Operations",
dated December 5, 1988, the licensee indicated that adequate toughness (i.e.,
Charpy USE less than 50 ft-lb) for the surveillance capsule specimens of controlling material existed through 32 EFPY (which corresponds to the design life of the plant).
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R~E UIREMENT.DOCUMENTS:
TITLE NUREG-0744, Revision 1, "Pressure Vessel tIaterial Fracture Toughness" Generic Letter 82-26, "Pressure Vessel Material Fracture Toughness" NUDOCS NO.
Cook Units 1 and 2
A<<11 DATE 10/82 11/12/82 2.
IMPLEMENTATION DOCUMENTS:
TITLE NUDOCS-NO.
DATE Alexich (AEP) to NRC 8812090036 12/5/88 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE
4 1
4
PLANT D. C. Cook -Units-1-and 2
DOCKET NO(S). 50-315.and 50-316 PROJECT MANAGER J. G.-Giitter TECHNICAL CONTACT R. Johnson
)RES)
USI NO.
A-12 TITLE Potential of Low.Fracture Tou hness and Lamellar Tearin
~n PMR SG an R
P Su orts MPA NO.
-A-07...
ISSUES
SUMMARY
TAC NOS.
This USI was resolved in October 1983 with the publication of NUREG-0577, "Potential of Low Fracture Toughness and Lamellar Tearing in Pl!R Steam Generator and Reactor Coolant Pump Supports."
The resolution contained no backfit requirements; it only applied to plants with a new construction permit issued after October 1983.
Standard Review Plan Section 5.3e4 was issued at the same time this USI was resolved.
The concern in this USI, as the title indicates, was the potential of low fracture toughness of some materials selected for fabrication of steam generator (SG) and reactor coolant pump (RCP) supports in operating PWRs.
Lamellar tearing was also of concern.
Fracture toughness is a measure of a material's resistance to fracture in the presence of a previously existing crack.
Generally, a material is considered to have adequate fracture toughness if it can withstand loading to its design limit in the presence of detectable flaws under stated conditions of stress and temperature.
The modifications to address this USI could involve maintaining minimum temperature around the supports above its fracture transition temperature, or total replacement of existing SG and RCP supports with supports fabricated of material grade which has a higher Charpy upper shelf energy and a lower transition temperature.
Analysis performed for the resolution of this USI determined that, even with the failure of the SG and RCP supports, the amount of incremental release of radioactivity would not be sufficiently high enough to justify any modification in terms of increasing the toughness of these supports.
This conclusion is based on a value-impact analysis documented in Appendix C of NUREG-0577.
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIFIC The licensee responded to the NRC's request for information concerning the fracture toughness of the steam generator and reactor coolant supports at D.
C.
Cook Units 1 and 2 in a letter dated November 23, 1977.
This information, initially reviewed by Sandia, was independently reviewed by Franklin Research Center.
Franklin reviewed the information per the criteria presented in NUREG-0577 and concluded the supports possess adequate fracture toughness.
The staff concurred with the this conclusion in a safety evaluation that was transmitted to the licensee on December 10, 1980.
REFERENCES:
1.
REQUIREMENT DOCUMENTS:
TITLE NUREG-0577, Rev. 1, "Potential of Low Fracture Toughness and Lamellar Tearing in PWR Steam Generator and Reactor Coolant Pump Supports" 2.
IMPLEMENTATION-DOCUMENTS:
NUDOCS-NO.
Cook Units 1 and 2
A-12 DATE 10/83 TITLE NRC to AEP NUDOCS NO.
8101080364 DATE 12/10/80 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE
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P LANT D. C.- Cook Units 1 and 2
DOCKET NO(S).
50-315 and 50-316 PROJECT MANAGER J. -G. Giitter TECHNICAL CONTACT D. Thatcher USI NO.
A-17 TITLE S stems Interactions in Nuclear Power Plants-.
MPA NO.
ISSUES
SUMMARY
TAC NOS.
Generic Letter (GL) 89-18, dated September 6, 1989, was sent to all power reactor licensees and constitutes the resolution of USI A-17.
The generic letter did not require any licensee actions.
GL 89-18 had two enclosures which (a) outlined the bases for the resolution of USI A-17, and (b) provided five general lessons learned from the review of the overall systems interaction issue.
The staff anticipated that licensees would review this information in other programs, such as the Individual Plant Examination (IPE) for Severe Accident Vulnerabilities.
Specifically, the staff expected that insights concerning water intrusion and flooding from internal sources, as described in the appendix to NUREG-1174, would be considered in the IPE program.
Also considered in the resolution of this USI was the expectation that licensees would continue to review information on events at operating nuclear power plants in accordance with the requirements of TMI Task Action Plan Item I.C.5 (NUREG-0737).
IMPLEMENTATION AND STATUS SUIUIARY (PLANT SPE~CIFIC:
AEP is currently reviewing Generic Letter 89-18.
The licensee stated in their November 29, 1989 response to Generic Letter 89-21 that systems interaction (e.g., internal flooding) will be addressed to some extent in their IPE program.
A brief description of the IPE program was provided to the NRC in a letter dated October 24, 1989.
A search of appropriate documents (i.e., Safety Evaluation and Supplements, FSAR, etc) for a potential licensee response or staff consideration of the September 1972 letter provided negative results.
Similarly, a preliminary search by the licensee did not yield a record of the September 1972 letter.
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REFERENCES:
1.
RE UIREMENT DOCUMENTS:
TITLE NUDOCS NO.
Cook Units 1 and 2
A-17 DATE Generic Letter 89-18 NUREG-1174 "Evaluation of Systems Interactions in Nuclear Power Plants" NUREG-1229 "Regulatory Analysis for Resolution of USI A-17" NUREG/CR-3922 "Survey and Evaluation of System Interaction Events and Sources" NUREG/CR-4261 "Assessment of System Interaction Experience in Nuclear Power Plants" NUREG/CR-4470 "Survey and Evaluation of Vital Instrumentation and Control Power Supply Events" HRC Letters to Licensees Informing Licensees of Staff Concerns Regarding Potential Failure of Non-Category I Equipment 2.
IMPLEMENTATION DOCUMENTS:
09/06/89 May 1989 August 1989 January 1985 June 1986 August 1986 9/72 TITLE NUDOCS NO.
DATE 3.
VERIFICATION DOCUMENTS:
TITLE NUDOC NO.
DATE
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PLANT D. C. Cook Units 1.and 2
PROJECT MANAGER J
G. Giitter DOCKET NO(S).
50-315 and-50-316 TECHNICAL CONTACT P.-Shemanski USI NO.
A-24 TITLE ualification-of Class 1E E ui ment MPA NO.
TAC NOS.
ISSUES
SUMMARY
This USI was resolved in July 1981 with the publication of NUREG-0588, Revision 1, "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment."
Part I of the report is the original NUREG-0588 that was issued for comment; that report, in conjunction with the Division of Operating Reactor (DOR) Guidelines, was endorsed by a Corrsoission Memorandum and Order as the interim position on this subject until "final" positions were established in rule making.
On January 21, 1983 the Commission amended 10 CFR 50.49 (the rule), effective February 22, 1983, to codify existing qualification methods in national standards, regulatory guides, and certain NRC publications, including NUREG-0588.
The rule is based on the DOR Guidelines and NUREG-0588.
These provide guidance on (a) how to establish environmental service conditions, (b) how to select methods which are considered appropriate for qualifying the equipment in different areas of the plant, and (c) such other areas as margin, aging, and documentation.
NUREG-0588 does not address all areas of qualification; it does supplement, in selected
- areas, the provisions of the 1971 and 1974 versions of IEEE Standard 323.
The rule recognizes previous qualification efforts completed as a result of Commission flemorandum and Order CLI-80-21 and also reflects different versions IEEE 323, dependent on the date of the construction permit Safety Evaluation Report (SER).
Therefore, plant-specific requirements may vary in accordance with the rule.
In summary, the resolution of A-24 is embodied in 10 CFR 50.49.
A measure of, whether each licensee has implemented the resolution of A-24 may therefore be found in the determination of compliance with 10 CFR 50.49.
This was addressed by 72 SERs for operating plants issued shortly after publication of the rule and subsequently in operating license reviews pursuant to Standard Review Plan Section 3.11.
This was further addressed by the first-round environmental qualification inspections conducted by the NRC.
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIFIC The staff issued orders dated August 29, 1980 (amended in September 1980) and October 24, 1980 to all licensees.
The August order required that the licensees provide a report, by November 1, 1980, 'documenting the qualification of safety-related electrical equipment.
The October order required the establishment of a central file location for the maintenance of all equipment qualification records.
The central file was mandated to be established by, December 1, 1980.
The staff subsequently issued a safety evaluation on,'nvironmental qualification of safety-related electrical equipment to the licensee on May 26, 1981.
The SER directed the licensee to "either provide
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documentation of the missing qualification information which demonstrates that safety-related equipment meets the DOR Guidelines or NUREG-0588 requirements or commit to a corrective action."
The information submitted by the licensee in response to the Nay 26, 1981 safety evaluation was evaluated for the staff by Franklin Research Center (FRC).
A safety evaluation transmitting the TER was sent to the licensee on January 17, 1983.
On February 22, 1983 the final rule (10 CFR 50.49) on environmental qualification of electrical equipment became effective.
On September 13, 1983, a meeting was held with the licensee to discuss resolution of environmental qualification deficiencies identified in the January 17, 1983 safety evaluation.
Subsequent submittals dated January 17, June 12, October 18 and December 10, 1984 documented the licensee's proposed method of resolving environmental qualification deficiencies and complying with 10 CFR 50.49.
On January 11, 1985, the staff issued a safety evaluation that concluded that the licensee's equipment qualification program is in compliance with 10 CFR 50.49 based on discussions during the September 13, 1983 meeting and review of the subsequent 1984 submittals.
In a letter dated January 25, 1985 the licensee stated under oath and affirmation
- "There is currently in place (at D. C. Cook) an Environmental Qualification (EQ) Program which is establishing the qualification requirements for the applicable electrical equipment.
lie believe this EQ program satisfies the requirements of 10 CFR 50.49 as we understand
- them, within the constraints currently approved by the NRC staff's Safety Evaluation (SE) dated January 11, 1985."
This letter went on to say that a request for an extension of the environmental qualification deadline was not anticipated.
However, in a letter dated June 28, 1985 the licensee requested an extension of the deadline to allow for the relocation of cables for the steam generator narrow range level differential pressure transmitters that were discovered to be routed below the maximum containment flood elevation.
The licensee committed to rerouting the cables for Unit 1 during the current outage (7/85) but requested an extension for Unit 2 until the next scheduled outage.
In a letter dated September 11, 1985 the staff informed the licensee that the request for an extension would not be considered.
In a [[letter::05000315/LER-1985-042, :on 850829,control Room Emergency Air Intake Damper HV-ACRDA-2 Discovered Improperly Positioned.Caused by Incorrect Damper Adjustment Made as Result of Surveillance Procedure 12 Thp 4030 STP.229.Damper Correctly Adjusted|letter dated September 30, 1985]] the licensee appealed to the Commission to consider an extension for Unit 2.
In a letter dated November 14, 1985, the Commission granted an extension until the next scheduled outage (no later than February 28, 1986).
The cables for Unit 2 were rerouted during this outage (3/86).
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REFERENCES:
1.
RE UIREMENT DOCUMENTS:
TITLE DOR "Guidelines for Evaluating Environmental gualification of Class 1E Electrical Equipment in Operating Reactors" NUREG-0588, "Interim Staff Position on Environmental qual ificat ion of Safety Related Electrical Equipment" Commission Memorandum and Order, CLI-80-21, on DOR Guidelines and NUREG-0588 NUREG-0588, Revision 1
10 CFR 50.49 (48 FR 2730-2733)
NUDOCS-NO.
Cook Units 1 and 2
A-24 DATE 12/79 05/23/80 07/81 01/21/83 Standard and Review Plan 3.11, Environmental qualification of Mechanical and Electrical Equipment 07/81 2.
IMPLEMENTATION DOCUMENTS:
TITLE Varga to Dolan (AEP)
Alexich (AEP) to Denton Alexich (AEP) to Palladino Chilk to Dolan (AEP)
NUDOCS NO.
8501290099 8501310460 8510070170 8511260216 DATE 1/11/85 1/25/85 9/30/85 11/14/85 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS-NO.
DATE
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Tl P
C I
J PLANT D. C. Cook Units 1 and 2
DOCKET NO(S).
50-315 and 50-316 JROJECT MANAGER J.
G.GTT TECRNTCAL CONT CT C~hLi USI NO.
A-26 TITLE Reactor Vessel-Pressure Transient Protection.-
MPA NO.
B-04 TAC NOS.
06805 and-06806 ISSUES
SUMMARY
This USI was resolved in September 1978 with the publication of NUREG-0224, "Reactor Vessel Pressure Transient Protection for PWRs,"
and Standard Review Plan Section 5.2.
The licensees of all operating PWRs were requested to provide an overpressure prevention system that could be used whenever the plants were in startup or shutdown conditions.
The issue affected all operating and future plants, and the staff established MPA 8-04 for implementing the solution at operating PHRs.
Since 1972, there have been numerous reported incidents of pressure transients in PWRs where technical specification pressure and temperature limits have been exceeded.
The majority of these events occurred while the reactors were in a solid-water condition during startup or shutdown and at relatively low reactor vessel temperatures.
Since the reactor vessels have less toughness at lower temperatures, they are more susceptible to brittle fracture under these conditions than at normal operating temperatures.
In light of the frequency of the reported transients and the associated potential for vessel
- damage, the NRC staff concluded that measures should be taken to minimize the number of future transients and reduce their severity.
Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its Impact on Plant Operations,"
was published July 12, 1988.
This generic letter provides guidance regarding review of pressure-temperature limits and indicates that licensees may have to revise low-temperature-overpressure protection (LTOP) setpoints.
IhlPLENENTATIDN AND-STATUS SUllllARY )PLANT.SPECIFIC):
By letter dated September 9, 1976 the licensee submitted to the NRC their initial response to the staff's August 13, 1976 request for a schedule and information to resolve the potential for overpressurizing the reactor coolant system.
A plant specific analysis in support of a reactor vessel overpressure mitigation system for Cook Unit 1 was submitted by the licensee in a letter dated August 4, 1977.
The staff concluded in a safety evaluation dated March 11, 1982, that the administrative controls and design modifications proposed by the licensee for Cook Unit 1 meets the criteria proposed by the NRC and is acceptable as a long term solution to the problem of overpressure transients.
The overpressure mitigation system (OMS) for Unit 2 was addressed in Supplement 7
to the D. C.
Cook Unit 2 Safety Evaluation Report issued in
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December of 1977.
The Supplement to the SER found that the ONS for Unit 2 was acceptable subject to the installation of equipment before twelve effective full power months of operation.
The licensee completed the installation within this time frame.
Technical Specifications covering the OHS for Unit 1 and Unit 2 were incorporated by Amendment 53 and Amendment 39, respectively.
On October 25, 1989 the licensee submitted revised pressure-temperature curves for Unit 2 based on recent surveillance capsule results and Regulatory Guide 1.99, Rev.
2 methodology.
Changes to LTOP setpoints were not necessary.
After analysis of surveillance capsule U is completed the licensee will submit pressure-temperature curves for Unit 1.
'E 1
REFERENCES:
l.
~RE UIREMENT DOCUMENTS:
TITLE
'NUREG-0224 - "Reactor Vessel Pressure Transient Protection for PWRs."
NUDOCS-NO.
Cook Units 1 and 2
A-26 DATE 9/78 NRC Letters to Licensees Informing Licensees of Staff Concerns Regarding Overpressure Low-Temperature Conditions in PWRs Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations" Standard Review Plan Section 5.2 August 1976 7/12/88 2.
IMPLEMENTATION DOCUMENTS:
TITLE AEP to NRC D. C. Cook Unit 2 SER, Supplement 7
3.
VERIFICATION DOCUMENTS:
NUDOCS NO.
DATE 8/4/77 12/77 TITLE NRC to AEP NUDOCS NO.
8204220602 DATE 3/11/82
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PLANT D. C.
Cook Units 1 and 2
DOCKET HO(S).
50-315 I and 50-316 PROJECT MANAGER J.
G. Giitter TECHNICAL CONTACT J. Wermiel USI HO.
A TITLE I I I E~L I PI I I MPA NO.
C-10 C-15 TAC NOS.
07980 07981 52217 and 52218 ISSUES
SUMMARY
This USI was resolved in July 1980 with the publication of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants,"
and Standard Review Plan (SRP) Section 9.1.5.
The staff established MPAs C-10 and C-15 for the implementation of Phases I and II, respectively, of the resolution of this issue at operating plants.
In nuclear power plants, heavy loads may be handled in several plant areas.
If these loads were to drop in certain locations in the plant, they may impact spent fuel, fuel in the core, or equipment that may be required to achieve safe shutdown and continue decay heat removal.
USI A-36 was established to systematically examine staff licensing criteria and the adequacy of measures in effect at operating plants, and to recommend necessary changes to ensure the safe handling of heavy loads.
The guidelines proposed in NUREG-0612 include definition of safe load paths, use of load handling procedures, training of crane operator s, guidelines on slings and special lifting devices, periodic inspection and maintenance for the crane, as well as various alternatives.
By Generic Letters dated December 22, 1980, and February 3, 1981 (Generic Letter 81-07), all utilities were requested to evaluate their plants against the guidance of HUREG-0612 and to provide their submittals in two parts:
Phase I (six month response) and Phase II (nine month response).
Phase I responses were to address Section 5.1.1 of NUREG-0612 which covered the following areas:
1 ~
2.
3.
4.
5.
6.
7.
Definition of safe load paths Development of load handling procedures Periodic inspection and testing of cranes gualifications, training and specified conduct of operators Special lifting devices should satisfy the guidelines of ANSI N14.6.6.
Lifting devices that are not specially designed should be installed and used in accordance with the guidelines of ANSI B30.9 Design of cranes to ANSI B30.2 or CMAA-70 Phase II responses were to address Sections 5.1.2 thru 5.1.6 of NUREG-0612 which covered the need for electrical interlocks/mechanical
- stops, or alternatively, single-failure-proof cranes or load drop analyses in the spent fuel pool area (PWR), containment building (PWR), reactor building (BWR),
other areas and the specific guidelines for single-failure-proof handling systems.
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As stated in Generic Letter 85-11, "Completion of Phase II of 'Control of Heavy Loads at Nuclear Power Plants' NUREG-0612," all licensees have completed the requirement to perform a review and submit a Phase I and a Phase II report.
Based on the improvements in heavy loads handling obtained from implementation of NUREG-0612 (Phase I), further action was not required to reduce the risks associated with the handling of heavy loads. Therefore, a
detai led Phase II review of heavy loads was not necessary and Phase II s(as considered completed.
While not a requirement, NRC encouraged the implementation of any actions identified in Phase II regarding the handling of heavy loads that were considered appropriate.
IIIPLENENTATION ANO STATUS SUNNARY (PLANT~SPECIFIC:
By letter dated December 22, 1980, the staff requested the licensee to review their provisions for handling and control of heavy loads at Cook Units 1 and 2
to determine the extent to which the guidelines of NUREG-0612 are presently satisfied.
The staff and its consultant, Franklin Research
- Center, reviewed the licensee's submittals and issued a safety evaluation on September 20, 1983.
The safety evaluation concluded that the guidelines in NUREG-0612, Sections 5.1.1 and 5.3 have been satisfied and, therefore, that Phase I for D.
C. Cook Units 1 and 2 is acceptable.
The licensee notified the staff by letter dated April 10, 1986 that modification of the drum pinion and gear of the main hoist to meet CHAA-70 (1975) standards had been completed in accordance with a previous commitment.
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RE UIREMENT DOCUMENTS:
TITLE Letter, Darrell G. Eisenhut, NRC, to all licensees, applicants for OLs arid holders of CPs transmitting NUREG-0612 and staff positions Generic Letter 85-11, Hugh L.
- Thompson, NRC, to all licensees for Operating Reactors, "Completion of Phase II of 'Control of Heavy Loads at Nuclear Power Plants'UREG-0612" 2.
IMPLEMENTATION DOCUMENTS:
NUDOCS HO.
Cook Units 1 and 2
A-36 DATE 12/22/80 06/28/85 TITLE Varga to Dolan (AEP)
Alexich (AEP) to Denton 3.
VERIFICATION-DOCUMENTS:
TITLE NUDOCS HO.
8310070408 8604160065 NUDOCS HO.
DATE 9/20/83 4/10/86 DATE
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I PLANT D.
C.
Cook Units 1 and 2
DOCKET NO(S).
50-315 and 50-316.
PROJECT MANAGER J.
G. Giitter TECHNICAL CONTACT A. Serkiz USI NO.
A-44 TITLE Station Blackout=
MPA NO.
A-22 TAC NOS.
68532 and 68533 ISSUES.
SUMMARY
This VSI was resolved in June 1988 with the publication of a new rule (10 CFR 50.63) and Regulatory Guide 1.155.
Station blackout means the loss of offsite ac power to the essential and nonessential electrical buses concurrent with turbine trip and the unavailability of the redundant onsite emergency ac power systems.
WASH-1400 showed that station blackout could be an important risk contributor, and operating experience has indicated that the reliability of ac power systems might be less than originally anticipated.
For these reasons station blackout was designated as a USI in 1980.
A proposed rule was published for comment on March 21, 1986.
A final rule, 10 CFR 50.63, was published on June 21, 1988 and became effective on July 21, 1988.
Regulatory Guide 1.155 was issued at the same time as the rule and references an industry guidance
- document, NUMARC-8700.
In order to comply with the A-44 resolution, licensees will be required to:
maintain onsite e'mergency ac power supply reliability above a minimum level develop procedures and training for recovery from a station blackout determine the duration of a station blackout that the plant should be able to withstand use an alternate qualified ac power source, if available, to cope with a station blackout evaluate the plant's actual capability to withstand and recover from a station blackout backfit hardware modifications if necessary to improve coping ability Section 50.63(c)(l) of the rule required each licensee to submit a response including the results of a coping analysis within 270 days from issuance of an operating license or the effective date of the rule, whichever is later.
IMPLEMENTATION=AND STATUS
SUMMARY
(PLANT SPECIFIC:
The licensee responded to the rule in a letter dated April 14, 1989.
In their response the licensee stated that the coping duration categories (e.g.,
AC and
II 1
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Emergency AC power classification) proposed for Cook are consistent with the guidance of NUMARC 87-00.
The submittal also identified procedures that would be modified, as necessary, to meet the guidelines in NUMARC 87-00:
These include procedures that identify loads that can be stripped from the batteries to provide a four hour capacity and procedures necessary to ensure backup methods exist for air operated valves needed for decay heat removal.
The safety evaluation for D.
C.
Cook Units 1 and 2 is scheduled to be completed during the second quarter of 1991..
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REFERENCES 1.
RE UIREMENT DOCUMENTS:
TITLE 10 CFR 50.63, "Loss of All Alternating Current Power" Regulatory Guide 1.155, Station Blackout" NUDOCS-NO.
Cook Units 1 and 2
A-44 DATE 06/21/88 08/88 2.
IMPLEMENTATION DOCUMENTS:
TITLE Alexich (AEP) to NRC 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
8904260177 NUDOCS NO.
DATE 4/14/89 DATE
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PLANT D C. Cook Units 1 and 2
DOCKET NO(S).
50-315 and 50-316.
PROJECT tlANAGER J.
G. Giitter TECHNICAL CONTACT P.
Y. Chen USI NO.
A-46 TITLE Seismic gualification of Equipment in Operating Plants HPA NO.
B-105.
TAC NOS.
ISSUES.
SUMMARY
USI A-46 was resolved with the issuance of GL 87-02 on February 19, 1987, which endor sed the approach of using the seismic and test experience data proposed by the Seismic qualification Utility Group (S(UG) and Electric Power Research Institute (EPRI).
This approach was endorsed by the Senior Seismic Review and Advisory Panel (SSRAP) and approved by the NRC staff.
The scope of the review was narrowed to equipment required to bring each affected plant to hot shutdown and maintain it there for a minimum of 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> s.
The review includes a walkthrough of each plant which is required to inspect equipment.
Evaluation of equipment will include:
(a) adequacy of equipment anchorage; (b) functional capability of essential relays; (c) outliers and deficiencies
( i.e., equipment with non-standard configurations);
and (d) seismic systems interation.
As an outgrowth of the Systematic Evaluation Program (SEP), the need was identified for reassessing design criteria and methods for the seismic quali-fication of mechanical equipment and electrical equipment.
Therefore, the seismic qualification of the equipment in operating plants must be reassessed to ensure the ability to bring the plant to a safe shutdown condition when subject to a seismic event.
The objective of this issue was to establish an explicit set of guidelines that could be used to judge the adequacy of the seismic qualification of mechanical and electrical equipment at operating plants in lieu of attempting to backfit current design criteria for new plants.
Generic Letter 87-02 with associated
- guidance, required all affected utilities to evaluate the seismic adequacy of their plants.
The specific requirements and approach for implementation are being developed jointly by S(UG and the staff on a generic basis before individual member utilities proceed with plant-specific implementation.
IMPLEMENTATION AND STATUS SUNf1ARY PLANT.SPECIFIC:
For All Plants:P1" i" d(G),R ii O.<<bi db HUG (of which AEP is a member) on June 3, 1988.
The staff issued a Generic Safety Evaluation (SE) on July 29, 1988 endorsing much of the GIP but with about 70 open items to be resolved.
After a series of meetings, SHRUG submitted Revision I to the GIP on December 23, 1988.
Supplemental information was
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REFERENCES:
1.
REQUIREMENT DOCUMENTS:
TITLE NUDOCS NO.
Cook Units 1 and 2
A-46 DATE Generic Letter 87-02, "Verification of Seismic Adequacy of Mechanical and Electric Equipment in Operating Reactors" 2/19/87 NUREG-1211, "Regulatory Analysis for Resolution of Unresolved Safety Issues A-46..."
NUREG-1030, "Seismic Qualification of Equipment in Operating Plants, Unresolved Safety Issue A-46" Letter attached with "Generic Safety Evaluation Report on SQUG GIP, Revision 0," from L. Shao (NRC) to
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Neil Smith (SQUG) 2.
IMPLEMENTATION DOCUMENTS:
02/87 02/87 07/29/88 TITLE NUDOCS NO.
DATE "Generic Implementation Procedure (GIP for Seismic Verification of Nuclear Plant Equipment," Revision 0
"Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment," Revision I 3.
VERIFICATION DOCUMENTS:
06/88 12/88 TITLE Alexich (AEP) to NRC NUDOCS-NO.
8810110222 DATE 10/3/88
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submitted by S(UG on March 17, 1989.
The staff has prepared a supplemental SE for GIP, Revision I and has submitted it to CRGR for review.
The target date for issuance of the supplemental SE is November 1989.
An additional supplement is scheduled for June 1990 and overall closeout of implementation projected for 1993.
Plant S ecific e
scensee responded to the NRC request for a plant-specific seismic verification plan for Cook by letter dated October 3, 1988.
In their letter the licensee stated that plant walkdown at Cook is expected to be completed by December 1992 provided that there are no major changes to the work scope and that the final SE (with no open items) is issued on schedule.
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REFERENCES:
1.
RE UIREMENT DOCUMENTS TITLE Generic Letter 89-19 "Request for Action Related to Resolution of USI A-47" NUREG-1217 "Evaluation of Safety Implications of Control Systems in LNR Nuclear Power Plants" HUREG-1218 "Regulatory Analysis for Resolution of USI A-47" 2.
IMPLEMENTATION DOCUMENTS:
NUDOCS NO.
Cook Units 1 and 2
A-47 DATE 09/20/89 June 1989 July 1989 TITLE NUDOCS HO.
DATE 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS HO.
DATE
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PLANT D. C. Cook Units 1 and 2
DOCKET NO(S).
50-315 and 50-316 PROJECT MANAGER J.
G. Giitter TECHNICAL CONTACT J. Kudrick USI NO.
A-48 MPA NO.
TITLE Hydrogen Control Measures and Effects of Hydrogen Burns on Safet Equipment TAC NOS.
49519 and 49520 ISSUES
SUMMARY
The NRC staff. concluded April 19, 1989, that USI A-48 is resolved, as stated in SECY 89-122.
USI A-48 was initiated as a result of the large amount of hydrogen gener ated and burned within containment during the Three Mile Island (TMI) accident.
This issue covers hydrogen'ontrol measures for recoverable degraded core accidents for all BklRs and those PNRs with ice condenser containments.
Extensive research in this area has led to significant revision of the Com-mission's hydrogen control regulations, given in 10 CFR 50.44, published December 2, 1981.
10 CFR 50.44 requires inerting of BWR Mark I and Mark II containments as a
method for hydrogen control.
The BWR Mark I and Mark II reactor containments have operated for a number of years with an inerted atmosphere (by addition of an inert gas, such as nitrogen) which effectively precludes combustion of any hydrogen generated.
USI A-48 with respect to BMR Mark I and II containments is not only resolved but understood to be fully implemented in the affected plants.
The rule for BllRs with Mark III containments and PHRs with ice condenser containments was published on January 25, 1985.
The rule required that these plants be provided with a means for controlling the quantity of hydrogen
- produced, but did not specify the control method.
In addition, the task action plan for USI A-48 provided for plant-specific reviews of lead plants for reactors with Mark III and ice condenser containments.
Sequoyah was chosen as the lead plant for ice condenser containments and Grand Gulf for Mark III containments.
Both of the lead plant licensees chose to install igniter-type systems which would burn the hydrogen before it reached threatening concentrations within the containment.
Final design igniter systems have been installed not only in both lead plants, Sequoyah and Grand Gulf, but in all other ice condenser and Mark III plants as well.
The staff's safety evaluations of the final analyses required to be submitted by these licensees by the rule are scheduled for completion in 1989.
Large dry PllR containments were excluded from USI A-48 because they have a
greater ability to accommodate the large quantities of hydrogen associated with a recoverable degraded core accident than the smaller Mark I, II, III and ice condenser containments.
However, this issue has continued to be considered and, in 1989, hydrogen control for large dry PNR containments was identified as a high-priority Generic Issue (GI) 121.
The resolution of GI 121 is being actively pursued in close coordination with more recent research findings.
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DOCKET NO(S).
50-315 and 50-316 PROJECT MANAGER J.
G. Giitter TECHNICAL CONTACT J.
Mauck US I NO.
A-47 TITLE Safety Implication of Control Systems in LWR Nuclear Power Plants MPA NO.
ISSUES
SUMMARY
TAC NOS.
79430.and 79431 USI A-47 was resolved September 20, 1989, with the publication of Generic Letter (GL) 88-19.
The generic letter states:
"The staff has concluded that all PWR plants should provide automatic steam generator overfill protection, all BWR plants should provide automatic reactor vessel overfill protection, and that plant procedures and technical specifications for all plants should include provisions to verify periodically the operability of the overfill protection and to assure that automatic overfill protection is available to mitigate main feedwater overfeed events during reactor power operation.
Also, the system design and setpoints should be selected with the objective of minimizing inadvertent trips of the main feedwater system during plant startup, normal operation, and protection system surveillance.
The Technical Specifications recommendations are consistent with the criteria and the risk considerations of the Commission.Interim Policy Statement on Technical Specification Improvement.
In addition, the staff recommends that all BWR recipients reassess and modify, if needed, their operating procedures and operator training to assure that the operators can mitigate reactor vessel overfill events that may occur via the condensate booster pumps during reduced system pressure operation."
Also, page 2 of the generic letter provides for additional actions for CE and BSW plants.
The generic letter provides amplifying guidance for licensees.
The generic letter requires that licensees provide NRC with their schedule and commitments within 180 days of the letter's date.
The implementation schedule for actions on which commitments are made should be prior to startup after the first refueling outage, but no later than the second refueling outage, beginning 9 months after receipt of the letter.
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIFIC:
The licensee is currently evaluating this USI.
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RE UIREMENTS
SUMMARY
CONT.):
Cook Units I and 2
A-48 The NRC staff has concluded that USI-A-48 is resolved as stated in SECY 89-122.
If interested, the report should be consulted for further details regarding the relationship of A-48 to other ongoing hydrogen activities.
IMPLEMENTATION AND STATUS
SUMMARY
(PLANT SPECIFIC:
RUREG-1370 Resolution. of Unresolved.Safety Issue A-AG, concluded, "an a aqua e
asks ex>sts Vonr ydrogen contro1 measures For degraded core accidents and that no new regulatory guidance for such accidents is necessary.
Therefore, the staff concludes that USI A-48 is resolved".
- However, NUREG-1370 also notes that the staff has required utility owners of ice-condenser containments to perform additional analyses to demonstrate equipment survivability for a broad spectrum of degraded core accidents and that these efforts (including staff review) will be completed in 1990.
Several additional hydrogen control issues have not been resolved for D. C.
Cook.
The licensee currently plans to address these issues as part of their Individual Plant Evaluation effort.
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REFERENCES:
1.
RE UIREMENT-DOCUMENTS:
. TITLE 10 CFR 50.44, Standards for Combustible Gas System in Light-)later-Cooled Power Reactors NUDOCS.NO.
DATE 12/81 SECY-89-122, Resolution of USI A-48, "Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment" 2.
IMPLEMENTATION DOCUMENTS:
04/19/89 TITLE NUDOCS-NO.
DATE 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS NO.
DATE
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PLANT D. C.
Cook Units 1 and 2
DOCKET NO(S).
50-315 and 50-316 PROJECT MANAGER J.
G. Giitter TECHNICAL CONTACT B. Elliott US I NO.
A-49 TITLE Pressurized Thermal Shock MPA NO.
A-21 TAC NOS.
ISSUES
SUMMARY
The final rule (10 CFR 50.61) on pressurized thermal shock (PTS) was approved by the Commission in July 1985.
, "Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for PWRs,"
was later published in February 1987.
Thus, this issue was resolved and new requirements were established, applicable to PHRs only.
The rule required that each operating reactor meet the screening criteria provided in the rule or provide supplemental analysis to demonstrate that PTS is not a concern for the facility.
Neutron irradiation of reactor pressure vessel weld and plate materials decreases the fracture toughness of the materials.
The fracture toughness sensitivity to radiation-induced change is increased by the presence of certain materials such as copper.
Decreased fracture toughness makes it more likely that, if a severe overcooling event occurs followed by or concurrent with high vessel
- pressure, and if a small crack is present on the vessel's inner surface, that crack could grow to a size that might threaten vessel integrity.
Severe pressurized overcooling events are improbable since they require multiple failures and improper operator performance.
However, certain precursor events have happened that could have potentially threatened vessel integrity if additional failures had occurred and/or if the vessel had been more highly irradiated.
Therefore, the possibility of vessel failure due to a severe pressurized overcooling event cannot be ruled out.
IMPLEMENTATION AND STATUS
SUMMARY
PLANT SPECIF~IC:
By letters dated January 22, 1986 and February 27, 1987, the licensee submitted information on the reactor vessel material properties and the fast neutron fluence in the reactor vessel beltline region for Cook Units 1 and 2.
The staff reviewed this information and issued a safety evaluation on March 27, 1987 that concluded that the Reference Temperature for PTS (RT-PTS) for both units is well within the 10 CFR.61 screening criteria through the end of the current operating license.
The SER also stated that the licensee should submit a reevaluation of RT-PTS as required by 10 CFR 50.61 whenever core loadings, surveillance measurements, or other information indicate a
significant change in projected values.
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REFERENCES:
1.
REQUIREMENT DOCUMENTS:
TITLE NUDOCS NO.
Cook Units 1 and 2
A-49 DATE 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Requirements" Reg.
Guide 1.154, "Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for PWRs" SECY 82-465, "Pressurized Thermal Shock" SECY 83-288, "Proposed Pressurized Thermal Shock Rule" Regulatory Guide 1.154 "Format and Content of P lant-Speci fic Pressurized Thermal Shock Safety Analysis Repor ts for Pressurized Water Reactors" Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations" 7/85 1/89 11/23/82 07/15/83 02/87 7/12/88 2.
IMPLEMENTATION-DOCUMENTS:
TITLE Alexich (AEP) to NRC Youngblood to Dolan (AEP) 3.
VERIFICATION DOCUMENTS:
TITLE NUDOCS-NO.
8601280125 8704020263 NUDOCS NO.
DATE 1/22/86 3/27/87 DATE INDEX
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Enclosure 3
USI Data Base Printout
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-~ p4 Page No.
02/01/90 LISTING OF IHCOtlPLETE USI DATA FOR INPUT FROM PROJECT MANAGERS ISSUE NUMBER ISSUE DESCRIPTIVE NAME IHPLEHEHT IMPLEMENT LICENSEE COHHENT STAFF COHHENT DATE STATUS 1f PLANT NAME: D. C.
COOK 2 A-01 A-02 MATER HAMMER ASYHHETRIC BLOMDOMH LOADS ON REACTOR PRIMARY COOLANT SYSTEMS MESTIH6HOUSE STEAM GENERATOR TUBE INTEGRITY CE STEAN BE}1ERATOR TUBE INTEGRITY BN STEAM GEERATOR TUBE INTE6RITY A-04 A-05 HARK I SHORT-TERtl PROGRAM HARK I LONG-TERtl PROGRAM MARK 11 CONTAINMENT POOL DYNAMIC LOADS - LON6-TERH PROGRAH AIMS BMR FEEDMATER HOZZLE CRACKIH6 A-06 A-07 A-08 A-09 A-10 A-11 A-12 A-17 A-24 A-26 A-31 A-36 A-39 VALVE POOL DYNAMIC LOADS AHD TEHPERATURE LIMITS SEISMIC DES16N CRITERIA-SHORT-TERN PRO6RAH PIPE CRACKS IN BOILINB MATER REACTORS COHTAINHEHT EHER6EHCY SUMP PERFORMANCE STATION BLACKOUT SHUTDOMH DECAY HEAT REMOVAL REQUIREtlEHTS SEISMIC QUALIFICATIOH OF A-40 A-42 A-43 A-44 A-45 A-46 EQUIPtlEHT IN OPERATIHB PLANTS SAFETY 1}lPLICATIOHS OF CONTROL SYSTEHS A-47 A-48 HYDR06EN CONTROL MEASURES AND EFFECTS OF HYDROGEN BURNS ON SAFETY EQUIPHEHT A-49 PRESSURIZED THERMAL SHOCK REACTOR VESSEL MATERIALS TOUGHHESS FRACTURE TOUGHNESS OF STEAH 6ENERATOR AND REACTOR COOLANT PUMP SUPPORTS SYSTEMS IHTERACTION QUALIFICATION OF CLASS 1E SAFETY-RELATED EQUIPMEHT REACTOR VESSEL PRESSURE TRANSIENT PROTECTION RMR SHUTDDMH REQUIREtlENTS CONTROL OF HEAVY LOADS HEAR SPEHT FUEL DETERHIHATION OF SAFETY RELIEF / I NC I / NC / / HC H/A H/A I I H/A / I N/A I I N/A 02/28/89 C I / N/A I / HC / / N/A I I HC 03/31/86 C 12/31/77 C I I H/A 04/10/86 C I I H/A I / NC I I H/A I / NC 03/31/93 I / I NC I I I 03/31/90 E 12/31/90 I 01/22/86 C ICE COHDENSER LEAK BEFORE BREAK I}1FO OHLY CE PLANTS ONLY BN PLANTS OHLY HK I BMR ONLY HK I BMR ONLY HK Il BMR OHLY BMR ONLY CP AFTER 83 ONLY HO REQUIRE}lENTS LTOPS HEM PLANTS ONLY. SRP. 6L-85-11 EHDED BMR ONLY SUBSUHHED BY A-46 BMR ONLY INFO ONLY SER 3/31/91 SUBSUMED BY SEVERE ACC REQ UNDER DEVEL HEN REQUIREMENTS UNDER NRC REVIEM
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Page Ho. 02/05/90 LISTIH6 OF INCOMPLETE USI DATA FOR INPUT FROM PROJECT NANA6ERS ISSUE NUMBER ISSUE DESCRIPTIVE MANE IMPLEMENT INPLENENT LICENSEE COMMENT STAFF COMMENT DATE STATUS ff PLANT A-01 A-02 A-03 1-04 1-05 A-06 1-07 A-08 A-09 A-10 A-11 e A-22 A-17 A-24 A-26 A-31 A-36 1-39 A-40 A-42 1-43 A-44 1-45 A-46 A-47 I -48 1-49 NANEs D. CD COOK 1 MATER HANMER ASYMMETRIC BLOMDOMH LOADS OH REACTOR PRINARY COOLANT SYSTEMS MESTIH6HOUSE STEAN 6EHERATOR TUBE INTE6RITY CE STEAM 6EHERATOR TUBE INTEGRITY BN STEAN 6EHERATOR TUBE INTE6RITY NARK I SHORT-TERN PROGRAM NARK I LONG-TERN PR06RAN MARK ll CONTAINMENT POOL DYHANIC LOADS - LOHB-TERN PR06RAN AT¹S BMR FEEDMATER HOZZLE CRACKIHB REACTOR VESSEL NATERIALS TOUBHHESS FRACTURE TOU6HHESS OF STEAN GENERATOR AND REACTOR COOLAHT PUNP SUPPORTS SYSTENS INTERACTION QUALIFICATION OF CLASS 1E SAFETY-RELATED EQUIPMEHT REACTOR VESSEL PRESSURE TRANSIEHT PROTECTIOH RHR SHUTDOMH REQUIREMENTS COHTROL OF HEAVY LOADS HEAR SPENT FUEL DETERNIHATIOH OF SAFETY RELIEF VALVE POOL DYHANIC LOADS AHD TENPERATURE LINITS SEISNIC DES16H CRITERIA-SHORT-TERN PR06RAM PIPE CRACKS IN BOILIHB MATER REACTORS CONTAINMENT EMER6EHCY SUMP PERFORNANCE STATION BLACKOUT SHUTDOMH DECAY HEAT REMOVAL REQUIRENEHTS SEISMIC OUALIFICATIOH OF EQUIPMENT IN OPERATIH6 PLANTS SAFETY INPLICATIOHS OF CONTROL SYSTEMS HYDROGEN CONTROL MEASURES AHD EFFECTS OF HYDR06EN BURNS OH SAFETY EGUIPNEHT PRESSURIZED THERNAL SHOCK / I HC I I HC / I HC / / N/A I I H/A I I N/A I I H/A / I H/A 07/08/89 C I I H/A / I NC I I N/1 I / HC 07/31/85 C 08/04/77 C I / N/1 04/10/86 C I I H/A / I NC I I H/A / I HC 03/31/93 I I I HC I / I 03/31/90 E 12/31/90 I 01/22/86 C ICE COHDENSER LEAK BEFORE BREAK INFO ONLY CE PLANTS ONLY BN PLANTS ONLY NK I BMR ONLY NK I BMR OHLY NK II BMR OHLY BMR OHLY CP AFTER 83 OHLY NO REGUIRENENTS LTOPS HEM PLANTS ONLY. SRP. GL-85-11 EHDED BMR OHLY SUBSUMMED BY A-46 BMR ONLY INFO OHLY SER 3/31/91 SUBSUMED BY SEVERE ACC REQ UNDER DEVEL HEM REGUIRENEHTS UNDER NRC REVIEM
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SAIC-90/1038 POST-IMPLEMENTATION AUDIT REPORT FOR INDIANA MICHIGAN POWER COMPANY'S DONALD C. COOK NUCLEAR POWER PLANT UNITS 1 AND 2 SAFETY PARAMETER DISPLAY SYSTEM FEBRUARY 21 AND 22, 1990 TAC Nos. 61207 and 61208 March 20, 1990 ~C7 U17~'" Science ApplicationstntemationalCtxtxxation Prepared for: U.S. Nuc1ear Regu1atory Commission Washington, D.C. 20555 Contract NRC-03-87-029 Task Order No. 36 Post Office Box 13tU, 1710 Goodridge Dnve, McLean, Virginia 221QZ, PN) 82W3N
i+) gg TABLE OF CONTENTS Section Pacae
1.0 INTRODUCTION
2.0 ACKGROUND................................................. B 3.0 VALUATION ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ E
- 3. 1 Concise Display of Critical Plant Variables to Control Room Operators........................................
3.2 Located Convenient to Control Room Operators.......... 3.3 Continuous Display of Plant Safety Status Information. 3.4 Should Have a High Degree of Reliability.............. 3.5 Suitably Isolated from Electrical and Electronic Interference with Safety Systems...................... 3.6 Designed Incorporating Accepted Human Factors Engineering Principles................................ 3.7 Minimum Information Displayed Should be Sufficient to Determine Safety Status with Respect to Five Safety Functions. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3.8 Procedures and Operator Training Addressing Operator Actions With and Without SPDS......................... 4.0 ONCLUSIONS................................................ C REFERENCES........................................................... 11 ATTACHMENT 1 - Agenda ATTACHMENT 2 - List of Attendees ATTACHMENT 3 - Licensee's Presentation Materials..
POST-IMPLEMENTATION AUDIT REPORT FOR INDIANA MICHIGAN POWER COMPANY'S DONALD C. COOK NUCLEAR POWER PLANT UNITS 1 AND 2 SAFETY PARAMETER DISPLAY SYSTEM FEBRUARY 21 AND 22, 1990
- 1. 0 INTRODUCTION The purpose of this report is to document the findings of a Nuclear Regulatory Commission (NRC) post-implementation audit of Indiana Michigan Power Company's Donald C. Cook Nuclear Plant, Units 1 and 2, Safety Parameter Display System (SPDS).
The NRC team consisted of two staff members from NRC's Human Factors Assessment Branch and a Contractor from Science Applications International Corporation (SAIC). The purpose of the February 21 and 22, 1990 onsite audit was to assess the status of the SPDS with regard to eight NUREG-
- 0737, Supplement 1, requirements (Reference 1).
The audit agenda is provided in Attachment 1. The list of meeting attendees is provided in Attachment 2, and licensee presentation materials are provided in Attachment 3.
2.0 BACKGROUND
The principle purpose and function of the SPDS is to aid control room personnel in determining the safety status of the plant and in assessing whether abnormal conditions warrant corrective action by operators to avoid degradation of the core. The SPDS can be particularly important during anticipated transients and the initial phase of an accident. All holders of operating licenses and applicants must provide an SPDS in the control rooms of their plants. The NRC requirements for the SPDS are defined in NUREG-0737,'upplement 1. This document requires licensees and license applicants to prepare a written safety analysis report sufficient to assess the safety status of each identified function for a wide range of events, including symptoms of severe accidents. Licensees and applicants must also prepare an
implementation plan for the SPDS that contains schedules for design, development, installation, and full operation of the SPDS as well as a design verification and validation plan. The safety analysis report and the implementation plan were to be submitted to the NRC for staff review. The staff's review is published in a safety evaluation report. The SPDS requirements, as defined by NUREG-0737, Supplement 1, are: 1. Provide a concise display of critical plant variables to control room operators (NUREG-0737, Supplement 1, Paragraph
- 4. l.a).
2. Be located convenient to control room operators (NUREG-0737, Supplement 1, Paragraph
- 4. l.b).
3. Provide continuous display of plant safety status information (NUREG-0737, Supplement 1, Paragraph
- 4. l.b).
4. Have a high degree of reliability (NUREG-0737, Supplement 1, Paragraph
- 4. l.b).
5. Be suitably isolated from electrical or electronic interference with safety systems (NUREG-0737, Supplement 1, Paragraph
- 4. l.b).
6. Be designed to incorporate accepted human factors engineering ~ principles (NUREG-0737, Supplement 1, Paragraph
- 4. l.e).
7. Minimum information displayed shall be sufficient to determine plant safety status with respect to five safety functions (NUREG-0737, Supplement 1, Paragraph 4.'1.f): i. Reactivity control ii. Reactor core cooling and heat removal from the primary system iii. Reactor coolant system integrity iv. Radioactivity control v. Containment conditions.
8. Implementation of procedures and operator training addressing actions with and without SPDS (NUREG-0737, Supplement 1, Paragraph
- 4. l.c).
Guidance for evaluating the application of the above requirements is provided by Appendix A to Section 18.2 of NUREG-0800 (Reference 2) and other documents cited therein, particularly NUREG-0700 (Reference 3). On April 12, 1989, the NRC issued Generic Letter 89-06: "Task Action Plan Item I.D.2 - Safety Parameter Display System - 10 CFR Part 50.54(f)" (Reference
/ 4). The letter requested all licensees of operating plants to perform an assessment and evaluation of the implementation status of their safety parameter display system. In addition, NUREG-1342, "A Status Report Regarding Industry Implementation of Safety Parameter Display Systems" dated April 1989 (Reference 5), was included with the generic letter to assist licensees in implementing SPDS requirements. The licensee responded to Generic Letter 89-06 by letter dated August 1, 1989 (Reference 6).
- 3. 0 EVALUATION The evaluation results are provided below in order of the eight NUREG-0737, Supplement 1,
SPDS requirements.
- 3. 1 Concise Dis la of Critical Plant Variables to Control Room 0 erators The evaluation of concise display included a review of physical location of displayed information and technical information organization within the display screens.
The SPDS was displayed on five screens located near the shift supervisor's desk. The licensee defined the SPDS as the two top level iconic displays (see ) along with a new Eberline digital radiation monitoring system and a containment isolation status system. Additional lower level displays were available on four of the SPDS screens, but were not part of SPDS. The combined technical content of the three SPDS display systems was sufficient to assess the status of the five NUREG-0737, Supplement 1, safety functions.
At the time of the
- audit, two parts of the SPDS were undergoing modifications and were defined by the licensee as non-operational.
The "RAD" spoke on the iconics was non-operational because it was driven by Westinghouse instruments that were replaced by the new Eberline digital radiation monitoring system.
- Second, the new containment isolation status display system was not operational because it was undergoing final modification and testing.
These systems were scheduled for operation following the 1990 refuelling outages. It was the review team's judgment that the licensee will meet the NUREG-0737, Supplement 1, requirement for a concise display when the Eberline radiation monitoring and the digital containment isolation status systems are completed and declared operational. 3.2 Located Convenient to Control Room 0 erators The evaluation of SPDS workspace location included an assessment of how the SPDS displays and controls supported the operator's needs during emergency operations. This included a determination of who was, defined by the licensee, a user of SPDS. The SPDS components consisted of four screens and keyboards that were used to display the iconics, and the containment isolation valve status
- system, and one screen with keyboard for the Eberline radiation monitoring system.
The five display screens were set up near the shift supervisor's desks in front of the control room. All screens and keyboards were convenient for use by the shift supervisor or shift technical advisor during emergency operations. It was the review team's judgment that the licensee met the NUREG-0737, Supplement 1, requirement for a display convenient to control room operators. 3.3 Continuous Dis la of Plant Safet Status Information The team evaluated the SPDS to determine if it continuously displayed information about the five critical plant variables identified in NUREG-0737, Supplement l.
The licensee limited its definition of SPDS to the two (wide and narrow range) iconic displays, the containment isolation status
- display, and the Eberline radiation monitoring system display.
This definition excluded the lower level displays available on the SPDS screens. The wide range SPDS iconic screen was designed to be displayed automatically on a reactor trip signal. This iconic provided an overview status of all five NUREG-0737, Supplement I, safety functions. However, in emergency operations, it would be necessary for the operators to call up the lower level detail displays for information not available on the iconic displays. This would result in loss of the iconics. The licensee did not have an administrative or technical system designed to maintain a dedicated iconic display of plant status. It was the review team's judgment that the licensee did not meet the NUREG-
- 0737, Supplement I, requirement for a continuous display because lower level detail displays could replace the overview iconic displays.
3.4 Should Have a Hi h Oe ree of Reliabilit In order to be judged reliable, SPDS should have a high degree of hardware and software availability. In the Generic Letter 89-06 checklist, the licensee stated that over the preceding 12 months the SPDSs in both units had achieved greater than 99/ availability. During the audit the iconic
- displays, Eberline radiation monitoring systems and containment isolation status systems in both units were operating reliably.
It was the review team's judgment that the licensee met NUREG-0737, Supplement I, requirement for a high degree of reliability.
3.5 Suitabl Isolated from Electrical and Electronic Interference with Safet 'Sstems The licensee's electrical and electronic isolation was evaluated previously by the NRC and found to be acceptable in an SER dated August 12, 1985. 3s6 Desi ned Incor oratin Acce ted Human Factors En ineerin Princi les The review team evaluated the human factors aspects of SPDS in the control room, including the SPDS technical
- content, display formats, and workstation designs.
The technical content of the iconic displays was found to be generally acceptable. However, the review team did identify two plant-specific concerns associated with the iconic displays. First, the Unit 2 iconic spoke for narrow range low steam generator level was set at 17%, which was correct for Unit 1 but not for Unit 2. Each unit had its own operating characteristics. Unit 2 low steam generator level setpoint was 21% and Unit 1 was 17%. The Unit 2 SPDS setpoint should have been modeled on Unit 2 rather than Unit l.
- Second, the Unit 1 Technical Specifications for high T-Avg was changed in June 1989, but the SPDS was not modified to reflect the change.
The high Unit 1 T-Avg should have been changed from 571.8'F to 570.9'F. These discrepancies were not apparent to the operators because the numerical values for the high and low level alarm setpoints on each parameter spoke of the narrow and wide range iconic displays were not displayed. Some alarm setpoints were reactor trip values and others were either anticipatory values or system capacity values. The formats of the iconic displays were reviewed and found acceptable. The review team evaluated the formats for the containment isolation valve status system and Eberline Radiation Honitoring System and found them acceptable.
The workstation design included five display screens and keyboards arranged on the shift supervisor's desk. This was a relatively large number of screens and keyboards. However the licensee did make an effort to consolidate SPDS information in a central location. The review team found the workstation design acceptable. It was the review team's judgment that the licensee did not meet the NUREG-
- 0737, Supplement I,
requirement for a display incorporating accepted human factors principles because of the incorrect parameter setpoints discussed above. I 3.7 Minimum Information Dis la ed Should be Sufficient to Determine Plant Safet Status With Res ect to Five Safet Functions The SPDS parameters used to depict the five NUREG-0737, Supplement I, critical safety functions are listed below. NUREG-0737, Supplement I Safety Functions Parameters 1. Reactivity Control Neutron Flux x Source Range x Intermediate Range x Power Range x Other: Turbine Impulse Pressure 2. Reactor Core Cooling and Heat Removal from the Primary System x x x x RCS Level. Subcooling Margin Hot Leg Temperature Cold Leg Temperature Core Exit Thermocouples Steam Generator Level Steam Generator Pressure RHR Flow Other: PRZR Pressure RCS Tavg RCS Pressure
. NUREG-0737, Supplement I Safety Functions Parameters 3. RCS Integrity X X x RCS Pressure Cold Leg Temperature Containment Sump Level Steam Generator Level Other: PRZR Level Wide L Narrow Range Net Charging Flow a. RCP I seal return flow b. RCP 2 seal return flow c. RCP 3 seal return flow d. RCP 4 seal return flow e. Letdown flow Radiation Monitoring Containment Temperature maximum Containment Pressure RV Level 4. Radioactivity Control 5. Containment Conditions X X X X x X Stack Monitor(s) Steamline Radiation(s) Containment Radiation Other: Effluent Monitors Plant Unit Vent, Airborne, Steam Jet Air Ejector Noble Gas, Gland Seal, Condensor Exhaust Noble Gas, Hain Steam Relief for Atmosphere, (SG PORV Monitor) Gross
- Gamma, Steamline Radiation Containment Pressure Containment Isolation Containment Hydrogan Concentration Other:
Containment Sump Level inputs Containment temperature inputs Radiation Monitoring inputs Residual heat removal (RHR) system flow and containment hydrogen concentration were not included in licensee's list of SPDS parameters. The review team determined that RHR flow in gallons per minute (GPH) was available on lower level display 3RCI-RCS, (see Attachment 3). Containment hydrogen concentration was also available on a lower level screen and alarmed at 4% on an
annunciation tile. Both RHR flow and containment hydrogen could be displayed when needed. It was the review team's judgment that the licensee met the NUREG-0737, Supplement 1, requirement for minimum information sufficient to determine plant safety status with respect to five safety functions. 3.8 Procedures and 0 erator Trainin Addressin 0 erator Actions With and Without SPDS The SPDS training consisted of less than 2 hours classroom, with no periodic update training. Interviews with operators and shift technical advisors revealed that they did not have a working knowledge of detailed SPDS information such as high and low setpoints on the iconic spokes. It was the review team's judgment that the licensee did not meet the NUREG-
- 0737, Supplement 1, requirement for procedures and training addressing operator actions with and without SPDS.
- 4. 0 CONCLUSIONS The purpose of the NRC's February 21 and 22, 1990 post-implementation audit of Indiana Michigan Power Company's Donald C. Cook Nuclear Plant, Units 1 and 2, Safety Parameter Display System was to assess the status of the system with regard to the eight NUREG-0737, Supplement 1, requirements.
As a result of the audit, it was the review team's judgment that the licensee met five of the eight requirements. The licensee did not meet the requirement for continuous display of plant safety status information because the overview status provided by the iconic displays could be replaced with lower level displays. The requirement for incorporation of accepted human factors principles was not satisfied because two parameter alarm setpoints were incorrect. Also, the requirement that operators should be trained with SPDS was not met because operators did not have working
knowledge of the high and low level alarm setpoints for each parameter on the narrow and wide range iconic displays. The licensee committed to evaluate the NRC's concerns and respond. 10
REFERENCES 1. NUREG-0737, Supplement 1, Requirements for Emergency
Response
Capability, Generic Letter 82-33,
- NRC, December 7,
1982. 2. NUREG-0800, Standard Review Plan of Safety Analysis Report for Nuclear Power
- Plants, Section 18.2, Rev.
0, Safety Parameter Display System (SPDS), Appendix A to SRP Section 18.2,
- NRC, November 1984.
3. NUREG-0700, Guidelines for Control Room Design
- Reviews, NRC, September 1981.
4. Generic Letter 89-06: Task Action Plan Item I.D.2 - Safety Parameter Display System - 10 CFR 50.54(f), NRC, April 12, 1989. 5. NUREG-1342, A Status Report Regarding Industry Implementation of Safety Parameter
- Systems, NRC, April 1989.
6. Generic Letter 89-06 Task Action Plan Item I.D.2 Safety Parameter Display System (SPDS)
- Response, Indian Michigan Power
- Company, August 1, 1989.
ATTACHMENT 1 AGENDA
TENTATIVE AGENDA FOR INDIANA HICHIGAN POh'ER COMPANY'S D.C. COOK - UNITS 1 AND 2 SAFETY PARAHETER DISPLAY SYSTEM AUDIT (February 21 through 23, 1990) l'ednesda Februar 21 1990 1:00 p.m. NRC Safety Parameter Display System (SPDS) Entrance Briefing SPDS Generic Letter 89-06 status Previous NRC findings regarding the D.C. Cook SPDS 1:15 p.m. Licensee presentation on SPDS 2:00 5:00 p.m. Control room and/or simulator orientation and begin SPDS discussion p.m. End Day 1 Thursda Februar 24 1990 8:30 a.m. SPDS discussion (continued) Completeness of parameters selected to dfsplay information about: l. 2. 3. 4. 5. Reactivity control Reactor core cooling and heat removal from the primary system Reactor coolant system integrity Radioactivity control Containment conditions 10:00 Generic Letter 89-06 (NUREG-1342)/D.C. Cook differences a.m. NUREG-0737, Supplement 1 requirements for: 1. Concise display of critical plant variables 2. Located convenient to control room operators 3. Should have a high degree of reliability 4. Continuous display of critical plant variables 5. Designed incorporating accepted human engineering prfncfples. 6. Procedures and operator training addressing actions with and without SPDS. 12:00 Lunch 1:00 p.th. NUREG-0737, Supplement 1 (continued) 2:00 p.m. NRC caucus 3:00 p.m. Open issues technical discussion wfth licensee/pre-exit brfeffng Frfda Februar 25 1990 8:30 a.m. Exit briefing
ATTACHMENT 2 LIST OF ATTENDEES
Attendees at the D.C. Cook - SPDS Audit Exit Meeting R. Correia C. Goodman J. DeBor B. Stoner J. Allard B. Jurgensen T. Kobetz J. Paris G. Hageniers R. Stephens R. B. Gridley W.T. Rae K.J. Toth NRC Technical Reviewer for SPDS Sr. Op. Engineer NRC Sr. H.F. Specialist Contractor (SAIC) Computer Science Computer Science NRC/Sr. Res. Insp. NRC/Reactor Engineer Computer Science, ACC Computer Science Operations NED/Engineer NOD/Engineer NOD/Licensing
ATTACHMENT 3 LICENSEE'S PRESENTATION MATERIAL
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