NUREG-0667, Forwards Request for Addl Info Re 860219,24 & 0312 Ltrs on 851226 Overcooling Transients & Corrective Actions.Startup Rept Should Include Info That NRC Identified as Necessary to Conduct Review

Forwards Request for Addl Info Re 860219,24 & 0312 Ltrs on 851226 Overcooling Transients & Corrective Actions.Startup Rept Should Include Info That NRC Identified as Necessary to Conduct Review

ML20206G471
Person / Time
Site:	Rancho Seco
Issue date:	05/30/1986
From:	Stolz J Office of Nuclear Reactor Regulation
To:	Julie Ward SACRAMENTO MUNICIPAL UTILITY DISTRICT
References
RTR-NUREG-0667, RTR-NUREG-0737, RTR-NUREG-667, RTR-NUREG-737, TASK-2.E.1.2, TASK-2.E.4.2, TASK-TM IEB-79-27, TAC-61635, NUDOCS 8606250315
Download: ML20206G471 (7)

Text

.. . - . ___

. .. May 30,1986 b Docket No. 50-312 Mr. John E. Ward Assistant General Manager, Nuclear (Acting)

Sacramento Municipal Utility District 6201 S Street P. O. Box 15830 Sacramento, California 95813

SUBJECT:

RANCHO SECO NUCLEAR GENERATING STATION - EFFECTS OF THE DECEMBER 26, 1985 OVERC00 LING EVENT

By letters dated February 19, 1986, February 24, 1986 and March 12, 1986, you submitted infornation on the December 26, 1985 overcooling transient and corrective actions you are taking as a result of the transient. Subsequently, you have enbarked on a management and performance improvement program which will include specific responses to the December 26, 1986 transient. As we understand it, your expanded restart report will be submitted on July 1, 1986. In the meantime, members of our technical staff have reviewed the already submitted information and found that additional information is required. We do not expect a separate response to cover this request for information. However, we suggest that your startup report include the information that the staff has identified as necessary to conduct its review.

The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore OMB clearance is not required under P.L.96-511.

Sincerely, ODIN MEW sIouangg Mgame.

John F. Stolz, Director PWR Project Directorate #6 Division of PWR Licensing-B

Enclosure:

Request for Additional Information cc w/ enclosure:

See next page

'; ,.IIISTR T RitT W ACRS-10 Q o_cket File BGrimes i

NRC PDR JPartlow L PDR SMiner 8606250315 860530 PBD-6 Rdg RIngram DR ADOCK 0500 2 g FMiraglia Gray File OELD EBrach EJordan H0rnstein 2a w PBD-6 PBD-6 -6 SMine RWeller 5/28/86 570/86 5/p/86

Mr. R. J. Rodriguez Rancho Seco Nuclear Generating Sacramento Municipal Utility District Station cc:

Mr. David S. Kaplan, Secretary Sacramento County and General Counsel Board of Supervisors Sacramento Municipal Utility 827 7th Street, Room 424 District Sacramento, California 95814 6201 S Street P. O. Box 15830 Ms. Helen Hubbard Sacramento, California 95813 P. O. Box 63 Sunol, California 94586 Thomas Baxter, Esq.

Shaw, Pittman, Potts & Trowbridge 1800 M Street, N.W.

Washington, D.C. 20036 Mr. Ron Columbo Sacramento Municipal Utility District Rancho Seco Nuclear Generating Station 4440 Twin Cities Road Herald, California 95638-9799 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 220, 7910 Woodmont Avenue Bethesda, Maryland 20814 Resident Inspector / Rancho Seco c/o U. S. N. R. C.

14410 Twin Cities Road Herald, California 95638 Regional Administrator, Region V U.S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek, California 94596 Director Energy Facilities Siting Division Energy Resources Conservation &

Development Commission 1516 - 9th Street Sacramento, California 95814 Mr. Joseph 0. Ward, Chief Radiological Health Branch State Department of Health Services 714 P Street, Office Building #8 Sacramento, California 95814 l

t

Enclosure REQUEST FOR ADDITIONAL INFORMATION RANCHO SEC0 NUCLEAR GENERATING STATION DOCKET NO. 50-312 The sections and pages listed below in Questions 1 through 21 refer to the February 19, 1986 submittal.

5 1.Section VII.1, page 42: Explain why the root cause of the loss of ICS power is tied to manufacturing error. The staff agrees that the root cause of the failure was an improperly installed lug, however these cabinets have been in service for a long time before this problem developed. If SMUD still believes the cause was at the manufacturing facility, has SMUD assured itself that the problem does not exist at other places in addition to ICS and NNI systems?

2.Section VI.l.b, page 38: Explain why SMUD cannot connect all the redundant power monitors (in two out of three arrangements) directly to the ,

d.c. source bus, rather than to the end of the distribution bus to avoid i

tripping all monitors with the same degraded signal. Also analyze the

" crosstalk" between the monitors if one trips to see if it creates any other failure mode.

3.Section V.3, page 15: Analyze the effect of partial loss of ICS power on different equipment to see that all essential equipment (controllers, indicators, etc.) that would be required for mitigating the consequences of this event is available to the operator.

4.Section V.3, page 15: Discuss the response of different equipment upon the restoration of power to ICS. How are the operators trained to handle the situation?

5.Section V.1.4, page 10: SMUD has provided an alternate pow q. supply for some equipment. However, insufficient informati,on has been provided to ascertain why other equipment that lost power as a result of this event should not be also powered from an alternate power supply. Provide the required information.

6.Section V.3.4.2.2 and V.3.4. 2.3, page 19: SMUD has stated that upon loss of ICS power, ADVs and TBVs will go to a " fail closed" position. The operator will be able to cycle these valves to maintain the steam pressure about a nominal value. SMUD has not provided any justification as to why this is the best alternative of controlling these valves. Also discuss how the valve position is indicated to the control room operator during the loss of ICS power.

7.Section V.I.4.b. page 11: Discuss alternatives to the present design response of the MFW start up valve, control valve and pump speed control on loss of ICS power and demonstrate why no change is required in the  ;

present design, particularly in view of the addition of feedwater to the l depressurized OTSGs by the condensate pumps which contributed to the overcooling / overfilling condition.

8. Section V.3.2, page 17: Provide the results of the independent laboratory testing of the power supply monitor and S1/S2 switches.

9. Section V.3.4.1, page 18: Explain how the display of the position of the 51/S2 switches would be changed so the operator can readily see whether the switches are in tripped condition.

i 10.Section V.4.1.2, page 21: Discuss the alarms and protective interlocks for the makeup /HPI pumps.

11. The present rate at which MSLFL is monitored by the Interim Data Acquisi-tion and Display System Computer (IDADS) is once per minute. However'during the event the pressure was decreasing approximately 70 psig/ minute. Hence, the MSLFL did not alarm until the pressure was much lower than the setpoint.

To resolve this concern SMUD has proposed a) a new setpoint and b) increase sample frequency for the MSLFL parameters by IDADS. However SMUD did not provide the new rate at which the parameters will be sampled. Provide the new rate and discuss why it is considered to be optimum.

12.Section V.1.9, page 13: Explain why it is not desirable to correct SPDS for OTSG level indication so that its readings reflect the actual condition rather than reading low.

13. Provide the basis for concluding that the plant satisfies the NUREG-0737; Item II.E-1.2, NUREG-0667 and IE Culletin 79-27 especially after the December 26, 1985 overcooling event. Provide the reference to the previous documents if they form the basis for such a determination. '

14.Section V.3.3, page 17: Explain the difference between the. casualty pro-cedures and the E0P. Howmuchattentionandtrainingisgitentothe operator for the casualty procedures?

15.Section V.3.4.2, page 18: Discuss if SMUD has considered a modes which could affect the ICS power as well as the alter *llaate . common failure power for the AFW control valves, TBVs and ADVs.

16.Section V.5, Page 21: Provide the results of the engineering evaluations' identified in this section concerning the damage to radiation monitor R-15001 which occurred during the December 26, 1985 event following. containment isolation after ESFAS actuation. Include a verification that loss of R-15001 did not affect the capability to achieve a safe shutdown. This confirmation should address the availability of other radiation monitoring instrumenta-tion and should also explain why the damaged radiation monitor does not affect the plant startup. Confirm that previous evaluations perfomed in connection with your response to Item II.E.4.2 of NUREG-0737 have properly identified those systems to be isolated 'following ESFAS actuation and that essential systems, those needed for safe shutdown or accident mitigation, j' do not receive a containment isolation signal following ESFAS actuation.

Also, provide a discussion of the effects of containment isolation on systems which are not initially required to be operable but are subsequently needed following ESFAS actuation. Confirm that isolation of these systems ,

does not result in adverse consequences.

i

_ _ - - _ _ - _ , , ~

1 1

17.Section VI.3, page 40: Explain in detail why it is acceptable to startup 4 the plant without the EFIC system even after the December 26, 1985 overcooling event?

18. Has SMUD taken credit for any non-safety system for mitigating the con-sequences of design basis events?

19. Provide a failure mode and effect analysis of ICS with interfacing systems to determine single failure points and their effects upon the loss of ICS power.

4

20. In accordance with NUREG-1195, the operators who investigated the loss of ICS power did not adequately understand the ICS power system configura-tion. Simplified drawings of the non-nuclear instrumentation (NNI) power l

supplies are posted on the NNI cabinets so the operator could benefit from these drawings quickly. Evaluate the desirability of having alse simplified electrical schematics of the ICS power supplies and ICS cabinets on the ICS cabinets.

21. Provide the basis for selection of the 280 gpm per OTSG flowrate as the preset failure position for the AFW flow control valves on loss of ICS power. Verify that this value does not result in an unacceptable under-cooling or overcooling assuming no operator action to control AFW flow for 10 minutes following a reactor trip from full power.

l 4

1 e

t

f

22. Provide a complete set of the latest Rancho Seco E0Ps.

23. a. Demonstrate that your procedures and training are adequate to respond to any potential failure involving NNI or the ICS (including but not limited to, overcooling events). The

! results of your ICS FMEA should be drawn upon in this activity. For all identified failure modes and plant responses, specifically identify the training elements and procedural steps which provide guidance to the operators in responding to the event. Show that these are adequate to minimize plant upset and enhance recovery a

following an ICS or NNI failure, and to ensure prompt and correct operator action. As a consequence of the December event, identify recent changes in procedures and training which will improve operator performance in a situation similar to the December event.

b. Consider and discuss (in greater detail than your February 19i.[1_906 letter),

the benefits achievable through development of event specific procedures related to ICS and NNI failures. Additionally, it is our understanding that a specific procedure dealing with loss of ICS power has been incorporated into the AT0G. Why was this not incorporated into the Rancho Seco E0Ps? Do you plan to incorporate this procedure into your E0Ps?

c. If a full range of ICS/NNI failures are not currently modeled on the plant simulator utilized by SMUD, provide a program and schedule for incorporating these elements.

24. In your February 19 memo, " Resolution of Issues Regarding the December 26, 1985 Reactor Trip", you indicate that an actual PTS problem did not occur. Yet, during the transient, PTS operational limits were exceeded.

The staff is concerned that a future overcooling event may again result in violation of PTS procedural guidelines, when the overcooling is combined with continuing high pressure injection and repressurization.

-- = _ _ . , -

The B&W AT0Gs heavily emphasize the maintaining of pressurizer level with HPI flow during a severe overcooling event. This is discussed in Chapter C, of Part II, Volume 1 dealing with the fundamentals of Reactor Control for Abnormal Transients. Little guidance seems to be provided in the consideration of vessel pressure limits during this event. Some balance between these potentially competing consideratiens may be in order.

Discuss the adequacy of the AT0G in providing guidance in this area. In areas where weaknesses are identified, demonstrate that your plant specific procedures are adequate to provide the necessary guidance to prevent further PTS violations during a similar event. If a revision to the PTS operational limits is advisable, present your approach and schedule for doing so.

You must also demonstrate that for the most severe overcooling transient possible, performance of the applicable E0Ps will not result in violation of PTS procedural guidelines. Document your identification and selection a of the most severe overcooling event, and discuss any procedural modifications which will be necessary to protect, PTS limits. ._ '

25. Provide the technical basis for your PTS Operational Limits. Describe your program for keeping abreast and evaluating modifications made at other facilities to improve plant safety, and whether the same modifications would improve plant safety at Rancho Seco.

26. Describe which position provided the control room command function during the course of the December 26, 1985 event.

i

27. In addressing the RCS overcooling transient (Item V.6 of the February 19, 1986 submittal), the conclusion is made that, "...the reactor vessel was not subjected to pressurized thermal shock as a result of this transient." The staff does not have sufficient information to agree with this conclusion. Therefore, we require information demonstrating that the reactor vessel beltline region has adequate structural integrity for the unit to return to service. Include in the information submitted a comparison of the December 26, 1985 transient cooldown rate, temperature and pressure with conditions that would cause crack propagation in the vessel. As they pertain to the information requested, the results of the B&W evaluation and EPRI analysis should be included in your response.

. - . - . . - - . - _.