ML19339B653
| ML19339B653 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 10/29/1980 |
| From: | Parker W DUKE POWER CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8011070386 | |
| Download: ML19339B653 (9) | |
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DUKE POWER COMPANY Powru Bustomo 422 Socin Cnuncu Srazzi, Cn2HLorTz. N. C. asa4a wnu4u o. pan =ca.sn.
October 29, 1980 vics Fatsiorms it L t e-ou t:Aacs7o4 Svtaas Paoovevion 373-4083 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission
'g Washington, D. C.
20555 n
d Attention:
Mr. B. J. Youngblood
(,
i Licensing Projects Branch No. I p;i g
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Subject:
McGuire Nuclear Station Docket Nos. 50-369 and 50-370 o
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Dear Mr. Denton:
i Enclosed with this letter are forty copies of an updated response to the document " Duke Power Company, McGuire Nuclear Station, Response to TMI Concerns." This document was transmitted to the NRC via my letter of May 23, 1980 and updated via my letters of July 18, 1980, August 6, 1980 September 8, 1980 and October 10, 1980.
Included in this review are changes to the McGuire Station Safety Engineering 4
Croup and changes to the Safety Evaluation of the Low-Power Test Program.
Ve'r/y truly yours, j
W ?
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William O. Parker, Jr.
TilH:scs Enclosures (40) 4 h
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s Mr. Harold R. Denton, Director October 29, 1980 Page Two WILLIAM O. PARKER, JR., being duly sworn, states that he is a Senior Vice President of Duke Power Company; and he is authorized on the part of said Company to sign and file with the Nulcear Regulatory Commission this docu-ment, Duke Power Company, McGuire Nuclear Station Response to TMI Concerns, and that all statements and matters set forth therein are true and correct to t e best of his, knowledge.
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/ William O. Parker, Jr.,~Vicd President Subscribed and sworn to before me this 29th day of October, 1980.
cL e.aRAo' Notary Public My Conmission Expires:
September 20, 1984
4 DUKE POWER COMPANY a
MCGUIRE NUCLEAR STATION j
Response to TMI. Concerns i
.l Changes and Corrections Remove These Pages Insert These Pages I-3A 10/10/80 I-3A 10/29/80-Appendix A Appendix A j
Station Directive 3.1.32 Station Direceive 3.1.32 (10/10/80)
(10/29/80)
Appendix E Appendix E 1-Page 3-4 Page 3-4 Page 3-5 Page 3-5 l
Page 4-13 Page 4-13 l
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The event investigator prepares a report describing the cause of the event and The
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any relevant plant behavior, and outlining proposed corrective actions.
SSEG then reviews each report.for occuracy and completeness, and assesses the s
adequacy of preposed corrective actions. The SSEG submits its report to the Station Manager and the NSR3 for review and for approval of corrective actions, to the Shift Technical Advisor for review with regard to operating procedures, and to the Supervisor of Training for inclusion of relevant information in the training program.
The SSEG will be composed of four full time engineers assigned to the day shif t.
The group will be staffed on a rotating basis frcm among experienced station personnel and will be multidisciplined with expertise in the areas of instrumen-tation, maintenance, operations, and technical services. Additional information on the membership and duties of the SSEG is provided in Station Directive 3.1.32 which is included in Appendix A.
Upon notification of an event, the General Office PC&L Section notifies company management, and may alert other General Office engineering and scientific support groups. The Station Manager forwards the approved event report to the General Office PCSL Section, where a Licensee Event Report (LER) is prepared and sub-mitted, if necessary.
Information is provided to other organizations, including NSAC, NRC, and the NSSS vendor. Detailed evaluations of plant transients are performed, and event occurrence data is maintained. As appropriate, other engineering support groups review the LER and station event reports for further recommendations on corrective actions, and may interface with appropriate equipment vendors. The NSRB performs an independent review of the event report, the LER, and the effectiveness of any follow-up actions.
O V
INDUSTRY EXPERIENCE EVALUATION Figure 2 illustrates the flow path for information received concerning industry operating experience. Significant events will be brought to the attention of Informa-Duke Power Company by NSAC, NSSS vendors, other utilities, or the NRC.
tion is distributed, as appropriate, to General Office engineering support groups for review and development of corrective actions and to the Training Services group for incorporation into the training program. The SSEG reviews the informa-s tion for applicability to the specific station, and makes recommendations to the NSR3 and the Station Manager in areas where action may be necessary. The Station Manager then developes and Laplements appropriate corrective actions with assistance from and review by the engineering support groups.
1-3A 10/."dSO J
STATION DIRECTIVE 3.1.32
) RAFT APPROVAL:
DATE:
FOR INFORMATION V
ANC/OR REVIEW GNLY DUKE POWER COMPANY MCGUIRE NUCLEAR STATION STATION SAFETY ENGINEERING GROUP OBJECTIVE This directive will define and establish administrative guidelines for function-ing of the on-site Station Safety Engineering Group (SSEG). The responsibilities and distribution of caporting for the SSEG will be outlinel below.
APPLICABP.ITY
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The SSEG will function as a full-time group consisting of four members and a permanent chairman. The Chairman of the SSEG will be appointed by and will f
report directly to the Director of the NSRE. All SSEG recommendations will be sent directly to the Director of the NSRB and the Station Manager. They in turn will assure proper action is taken to address SSEG recommendations.
MEMBERSHIP J
'V All members of the SSEG shall have at least six years of technical experience with a minimum of two years being nuclear station experience. A maximum of four years of the six years may be fulfilled by academic or related technical training. The four members of the SSEG will be chosen from the attached list (Attachment 1) with at least one person from each of the four areas listed.
These members will be selected by and will report to the Chairman of the SSEG during their assignment in this area. Personnel will be assigned to this area for at least six months.
RESPONSIBILITIES The SSEG will function as an independent technical reviev group in the areas outlined below.
A.
Licensee Event Report (LER) Summaries - Each month the General Of fice Licensing Group will forward a su= mary list of all LER's applicable to McGuire to the SSEG. The Chairman will assign these LER's to members of the SSEG to determine if any specific action needs to be taken at McGuire to prevent or mitigate the consequences of a similar event. All recommendations will be forwarded directly to the responsible station group, the Station Manager and the Director of tha NSR3.
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B.
Effectiveness of Plant Programs - The Chairman will assign members of the SSEG specific station programs to review and O,
determine if any recommendations to increase the effectivcness of the program should be considered. After review by each member of the SSEG, any recommendations will be forwarded to the responsible station groups, the Station Manager and the Director of the NSRB.
C.
Plant Modification Review - All design changes involving struc-tures, systems, or components with QA conditions will be reviewed by the SSEG. This review is to insure all safety concerns are properly addressed.
D.
Station Procedures and Changes - Selected station procedures and/or changes to procedures will be reviewed by the SSEG to determine their adequacy.
E.
Plant Indicent Reports - All incidents involving reportable items as defined by Station Directive 2.8.1 or other investigations as deemed appropriate by the Chairman of the SSEG will be assigned to a member of the SSEG for investigation and preparation of the station incident report as outlined in Station Directive 2.8.1.
The incident report will be reviewed by the SSEG and any additional recommendations included. These reports will be sent to the Station Manager, the Director of the NSRB, and other groups as outlined by Station Directive 2.8.1.
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inoperable by forcing the logic to see that the reactor trip breakers
/N are open. Westinghouse believes 'that this mode of operation is accep-t)
table for the short period of time these tests will be carried out based on the following:
1.
Close observation of the partial trip indication by the operator, Rigid adherence to the operator action points as defined by West-2.
inghouse, see Section 3.2.
3.
Little or no decay heat is present in the system, thus Safety Injection serves primarily as a pressurization function.
(Shutdown Margin cap 4-bility is considerably more than 1.6% SK/K for control rods at or above insertion limits).
Stocking these functions will allow the performance of these tests at icw power, pressure, or temperature and close operator surveillance will assure initiation of Safety Injection, if required, within a short time period.
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Lowering the automatic auxiliary feedwater start will have little effect, since there is lictie or no decay heat present. Close operator surveillance will insure auxiliary feedwater addition if necessary.
3.1. 7 T. S. 3.4.4 PRESSURIZER The Pressurizer provides the means of maintaining pressure control for the plant. Normally this is accomplished through the use of pressurizer heaters and spray. In several tests the pressurizer heaters will be either turned off or rendered inoperable by loss of pcwer. This mode of operation i acceptable in that pressure control will.be maintained through the use of pressurizer level and charging / letdown flew.
3.1.8 T.S. 3.7.1.2 AUXILIARY FEEDWATER SYSTEM The auxiliary feedwater system will be rendered partially inoperable f or two tests. The two t e s t s simula t e s eme f orm o f los s o r AC power, i.e.,
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notor drivan auxiliary feodwetor pumps inoperable. Westinghouse has determined diat diis is acceptable f or these two tests because of the little or ne decay heat present allowing sufficient time (# 30 min-utes) for operating personnel to rack in t5e pump power supplies and regain steam generator level.
3.1.9 T.S. 3.3.1.1, 3.8.2.1, 3.3.2.3 POWER SO'JRCES
- hese specifications are outside Westinghouse control, however it is acceptable to alter power source availability as long as manual Safety Injection is operable and safety related equipment will function Wien required.
3.1.10 T.S. 3.10.3 SPECIAL TEST EXCEPTIONS - PHYSICS IESTS This specification allows the minimum temperature f or criticality to be as low as $41*F.
Since it is expected that RCS T will be taken avg as low as 085 F esis specification will be excepted. See Section 3.1.4 for basis of acceptability.
3.1.I1 TECHNICAL SPECIFICATIONS NOT EXCEPTED I
While not applicable at power levels below 5: RTP the following tech-nical specification limits can be expected to be exceeded:
1.
3.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z) q At low temperatures and flows F (2) can be expected to be above q
normal for 5% RTP with RCPs running. However at such a low power level no significant deviations in burnup or Ie peaks are expected.
2.
3.2.3 RCS FLOW RATE A20 R At low temperatures and flow Fgg can be expected to be higner dian if pumps are running. However, no significant consequences for full power operation are expected.
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i the Analysas of coro conditions bassd on th2so essemptions indicate that DN3 criterion of the FSAR is met.
s 4.2.3.3 Secondarv Pressure Trip Protection Large steamline ruptures which affect all loops uniformly will actuate reactor trip and steamline isolation on Law Steamline Pressure signals Low Pressurizer Pressure and Power Range Neutron Flux in any ona loop.
icw setpoint trips serve as further backups. An example is the double-ended rupture of a main steamline downstream of the isolation valves, with all isolation valves initially open. Figures !%?,7 2nd 4.2.8 show the response to such an event, wi:h an inizial power of 5%
and natural circulation. The Low Steamline Pressure trip occurs almost In the example shown, the main steamline isolation valve immediately.
on loop one was assumed to fail to close. No power excursion resul:ed, and the reactor remained subcritical af ter the trip.
4.3 ADDITIONAL CONSIDERATIONS
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In the great majority of cases it was concluded, either by reanalysis or fuel clad by comparisen with previously analyzed FSAR conditions, that integrity would be maintained without need for operator mitigating the accion. For the LOCA or steambreak events, it was concluded that operator would have more than ample time (> 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) to respond by manual accion, e.g., manually initiate saf ety injection, to preclude fuel damage.
Finally, in certain other cases, primarily associated with certain inadvertent RCCA withdrawal events, the postulated accident conditions were neither amenable to direct analysis nor credi: for operator inter-vention. In particular, the postulated accident conditions were outside the bounds of accepted analysis techniques so that fuel damage was not For chese precluded either by analysis or identified operator action.
cases, the basis f or acceptability was primarily associated with the low probability of an inadvertent rod withdrawal event during the limi:ed duration of the special :ests.
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