ML20203N882

Discusses NUREG-1195, Loss of Integrated Control Sys Power & Overcooling Transient at Rancho Seco on 851226. Minor Technical Corrections to & Comments on Misperceptions in Incident Investigation Team Rept Encl

ML20203N882
Person / Time
Site:	Rancho Seco
Issue date:	04/24/1986
From:	Reinaldo Rodriguez SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Miraglia F Office of Nuclear Reactor Regulation
References
RTR-NUREG-1195 RJR-86-168, TAC-60462, NUDOCS 8605060174
Download: ML20203N882 (5)

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SACRAMENTO MUNICIPAL UTILITY DISTRICT ' ' 6201 S Street. P.O Box 15830. Sacramento CA 95852 1830.(916)452 3211 AN ELECTRIC SYS T EM SERVING T HE HE AR T OF CAllf 0RNIA RJR 86-168 April 24, 1986 DIRECTOR OF NUCLEAR REACTOR REGULATION ATTN FRANK J MIRAGLIA JR DIRECTOR PWR-B DIVISION U S NUCLEAR REGULATORY COMMISSION WASHINGTON DC 20555 DOCKET NO. 50-312 RANCHO SECO NUCLEAR GENERATING STATION UNIT NUMBER 1 NUREG-1195: LOSS OF INTEGRATED CONTROL SYSTEM POWER AND OVERC00 LING TRANSIENT AT RANCHO SEC0 ON DECEMBER 26, 1985 The District has reviewed the report by the NRC Incident Investigation Team (IIT), NUREG-1195, and agrees with the technical descriptions of the Rancho Seco systems and their operation.

However, for complete-ness, minor technical corrections have been identified and are provided in.

The District would also like to take this opportunity to address what may be misperceptions included in the report.

These comments are provided as Attachment 2.

The District will be addressing the findings and conclusions of the IIT, t

as identified in NUREG-1195, in. future submittals.

If there are any questions concerning the attached material, please contact Mr. Robert Little at (916) 732-6021.

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.J.hDRIU ASSISTANT GENE L MANAGER, NUCLEAR

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Attachments 8605060174 860424 PDR ADOCK 05000312

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RANCHO SECO NUCLEAR GENERATING STATION O 14440 Twin Cities Road, Herald, CA 95638-9799;(209) 333 2935

ATTACHMENT 1 DISTRICT TECHNICAL COMMENTS ON NUREG-1195 NUREG-1195 NUREG-1195 STATEMENT REFERENCE (SECTION N0.)

DISTRICT COMMENTS High pressure steam from 3.3 Main Steam is used at low loads the Main Steam System and for single pump operation supplements low pressure at high loads. Low Pressure steam at higher plant Steam is adequate for two pumps loads. Steam to MFW at full load, which is the pumps is provided from normal operating configuration, auxiliary boiler at low power operations.

MFW control valves 3.3 Valves FV-20575 and FV-20576 (FV-20575 and FV-20576) are globe valves.

are gate valves.

AFW supplied through circu-3.4 AFW is supplied through an exter-lar header in steam annulus nal header with six nozzles on each steam generator.

f OTSG high level alarm is 3.4.2 Setpoint is 92.5 percent on the at 82.5 percent.

operate range.

Failed makeup pump had only 4.6 In addition to the mentioned a single stop-check valve stop-check valve, there is a that isolated RCS from failed check valve on the pump dis-makeup pump seals.

charge.

RCS low pressure was 1064 psig.

4.4 RCS low pressure actually 1046 psig.

Site boundary dose was 0.2 4.6 Site boundary dose calculated mrem.

6.9 to be 0.02 mrem.

80 curies were released.

6.9 Re-evaluation shows that 33 curies of noble gasses were released.

SFV-20577 (Aux. FW to OTSG-A) 9.2 Both SFV-20577 and SFV-20578 will not open if offsite AC are diesel backed with safety-power is not available.

grade controls,

m ATTACHMENT 2 DISTRICT COMMENTS ON NUREG-1195 NUREG-1195 NUREG-1195 STATEMENT REFERENCE (SECTION NO.)

DISTRICT COMMENTS SMUD personnel found doing 2.4 Immediately following the troubleshooting in a highly 10.2.9 10-2-85 reactor trip, the controlled, systematic, and District instituted a systematic well documented manner quite program for analyzing the event different than normal mainte-and resolving root causes.

The nance.

program was based up NUREG-ll54 Appendix B, which described the Davis Besse systematic trouble-shooting program. The District's program was again implemented following a trip on 12-2-85.

As a result, Region V approved re-start.

Following the 12-26-85 event, the District again implemented the sys-tematic troubleshooting program.

The program was in full effect when the IIT arrived on site.

The IIT performed a line by line comparison of the District's program with NUREG-1154 Appendix B without con-sidering procedural and organiza-tional differences between the two organizations.

As a result, the District's plan was revised twice to incorporate IIT wording which in the District's opinion did not constitute substantive changes affecting the outcome of the troubleshooting.

The District had a highly con-trolled, systematic, and well documented troubleshooting program in place prior to, and following the 12-26-85 event.

SMUD personnel had consid-2.4 The District's investigations erable difficulty in providing 10.2.10 following the 12-26-85 event detailed information wnich were broad in scope and exceeded delayed the team's investiga-that initially identified by the tion.

IIT. As the IIT increased their knowledge of the plant and the

ATTACHMENT 2 (Page 2)

DISTRICT COMMENTS ON NUREG-1195 NUREG-1195 NUREG-1195 STATEMENT REFERENCE (SECTION NO.)

DISTRICT COMMENTS event, they expanded their areas of interest. Often the District already had an investigation underway in the area of question.

This may have lead the IIT to per-ceive that the District was with-holding information. As stated in NUREG-1195, when requested, the detailed information was pro-vided.

The District did have difficulty in anticipating which areas the IIT would desire detailed information.

Significantly, the detailed infor-mation was available when requested indicating that the District had independently implemented an effec-tive troubleshooting program.

The IIT was slow to respond to effers to locate them in more effective work space, which had an adverse impact upon the inter-face with District personnel.

Operators did not diagnose 4.3 The operators responded in accord-Loss of ICS Power for the 10.2.1 ance with the symptom based E0Ps first 2 minutes of the immediately, and had determined incident the loss of ICS prior to the reactor

trip, i.e., within the first 15 secs.

Operators neither applied 6.5 This statement is not substantiated nor understood the signi-in the IIT Report or in the Dis-ficance of the E0P rules, trict's evaluation of the operators and their adherance to procedures.

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I ATTACHMENT 2 (Page 3)

DISTRICT COMMENTS ON NUREG-1195 i

NUREG-1195 NUREG-1195 STATEMENT REFERENCE (SECTION N0.)

DISTRICT COMMENTS i

NUREG-1195 implies that the 6.9 The Control Room was, and continues t

i Radiation Protection Techni-to be, responsible for the AP.509 cian was responsible for Emergency Plan Offsite Dose offsite exposure calculation Calculations.

per AP.509.

NRC st6ff was led to believe 7.2.4.3 The AFW/EFIC (II.E.1.2) scope and that EFIC would be installed 10.1.11 schedule changes had been provided in 1984 (II.E.1.2).

SMUD to the NRC staff.

installed alternate system and The NRC issued Safety Evaluation l

not make it clear to NRC Reports in Jan. and Sept. 1982, assuming EFIC installation.

In i

Oct. 1982, the District indicated i

that it would install interim safety grade AFW modifications and that EFIC was separate and beyond the AFW upgrade requirements of NUREG 0737. The District also sub-mitted a new schedule for EFIC i

implementation showing completion by Cycle 7. This schedule was con-firmed by the District in Dec. 1982.

In April 1983 the District submitted a revised AFW system description describing the interim AFW upgrades.

NRR confirmed their understanding 3

in an SER on the status of the AFW system dated Sept. 26, 1985.

In Nov. 1983 the District indicated that the interim system was installed 4

thus completing Item II.E.1.2.

Via a series of letters and living schedule submittals, the District i

informed the NRC that the EFIC I

installation was scheduled in two phases (Cycle 8 and Cycle 9).

l This understanding and approach i

was confirmed with NRR during a i

meeting in Oct. 1985, at which time, the District committed to install the bulk of EFIC during The Cycle 8 outage.

l NUREG-1195 infers that SMUD 7.2.4.3 AFW initiation on SFAS was in installed AFW initiation on place upon issuance of the SFAS as a modification, original operating license.

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