ML18081B124

Technical Evaluation Rept of Electrical,Instrument & Control Aspects of Inadvertent Safety Injections at Salem, Unit 1.

ML18081B124
Person / Time
Site:	Salem
Issue date:	01/31/1980
From:	Cleveland C EG&G, INC.
To:	Shemanski P Office of Nuclear Reactor Regulation
References
CON-FIN-A-6256 EGG-EA-5083, NUDOCS 8002280068
Download: ML18081B124 (10)

Text

n EGI:G e-'> Idaho, Inc.

FOAM EG&G-398 *

(Rev 12-78)

INTERIM REPORT Accession No. - - - . - - - - - - -

Report No. _E_G_G-_E_A_-_5_08_3_ __

Contract Program or Project

Title:

Electrical, Instrumentation and Control System Support Subject of this Document:

Technical Evaluation Report of the Electrical, Instrument, and Control Aspects of Inadvertent Safety Injections at Salem, Unit l (Docket 50-272)

Type of Document:

• Technical Evaluation Report Author(s):

C. J. Cleveland Date of Document:

January 1980 Responsible NRC Individual and NRC Office or Division:

Paul Shemanski, Division of Operating Reactors This document was prepared primarily for preliminary or internal use. It has not received full review and approval. Since there may be substantive changes, this document should not be considered final.

EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission and the U.S. Department of Energy Idaho Operations Office Under contract No. DE-AC07-76ID01570 NRC FIN No.

A6256 INTERIM REPORT

TECHNICAL EVALUATION REPORT ELECTRIC, INSTRUMENT, AND CONTROL ASPECTS OF INADVERTENT SAFETY INJECTIONS AT SALEM, UNIT 1 DOCKET 50-272 by C. J. Cleveland

CONTENTS J. 0 INTRODUCTION * . . . . .... ... . . .. 1

2.0 DESCRIPTION

AND EVALUATION OF SAFETY INJECTIONS 1 3.0 CON CL US ION * ........ 6 3.1 Salem Safety Injections . 6 3.2 Generic S!s for PWRs 7

4.0 REFERENCES

• . . . . . . ... 7

I TECHNICAL EVALUATION ELECTRIC,* INSTRUMENT, AND CONTROL ASPECTS OF INADVERTENT SAFETY INJECTIONS AT SALEM, UNIT 1

(

DOCKET 50-272 '

-1. 0 INTRODUCTION Since November 30, 1976~ the Public Service.Electric and Gas Com-pany's ( PSE&G) Salem Un it No. 1 Nucl ea~ St~ti on, a Westinghouse four (4) loop PWR, has experienced eleven (11) inadvertent safety injections (SI s).

A Westinghouse letter dated Decembar 13, 1976, to PSE&G states that the SI nozzlescan withstand* fifty. (SO) Sis*with a 40°Fwater transient before the appropriate stress. 1imi ts of the nozzles are exceeded ..

The objectives of th.is report are to:

(1) Review the electrical, instrument, and controls (EI&C) of the safety injection systems to determine if changes wi 11 be required to prevent future i nad-vertent safety injections.

• (2) . Determine if any generic problems exist in other PWRs concerning inadvertent Sis.

2.0 DESCRIP°TION AND EVALUATION OF SAFETY INJECTIONS.

A11 of the docketed Lken see Event Reports and Emergency Core Cooling System Actuation Reports were examined to obtain the following surrmary of the. events. and the-licensee'.s proposed corrective actions to prevent recurrence of these events:

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(a) Safety Injections 1, 2, and 3 were caused by the premature lifting of a safety relief valve on a steam generator. Contrary to information from the manufacturer, the valve actuation setting changed during maintenance. The licensee instigated admin-istrative controls such that in the future these type valves would be tested and adjusted as neces-sary after any maintenance actions that could

~ffect the set point. All other like valves were*

tested after SI No. 3.

(b) Safety Injection 4 was caused when a technician mistakenly used grounded test leads when hooking up a brush recorder to various steam fl ow and steam pressure channels. The licensee verified this cause during a special test within a few days after Event 4. To prevent this type of occurence from repeating, all brush recorders were modified-to allow use of grounded test leads. Technicians and their supervisors were advised of the modification.

(c). Event 5 was the result of an operator mistakenly setting the set point of a steam dump controller at 980 psig instead of the required 1005 psig. This setting in conjunction with a reactor trip opened the steam dump resulting in a high steam flow sig-nal. A review of the incident reports did not reveal any proposed corrective action by the licensee.

(d) Safety Injection 6 was caused by failure to follow proper maintenance procedures. Maintenance person-nel were repairing a hyrlraulic hose failure on No. 12MSIV without blocking the operation of the hydraulic valve motor. As fluid was added the valve cycled open resulting in a high steam line 2

differential pressure' signaL *Again, the *ECCS report did not cite any corrective action to reduce the chances of a reo.ccurrence of this type.

(e) Safety _Injection 7 resulted when an operator mis-takenly pushed th{wrong push-buttons while-per~

forming a surveillance procedure. The very close proximity of the controls was a contributing factor. No corrective action was given b~ the .

• licensee in the ECCS report.

(f) Events 8 and 9 were caused by a mis adjustment of a lead/lag controller in the steam dump system.*

Investigation after the latter incident by the licensee verified t_hat the output.of the controller was lagging the input, prohibiting the modulation' of the steam dump valves, thereby causing T ve to increase instead of being controlled at 5478F. A review of work orders and data cards* rev.ea 1ed the ,

cont~oller was calibrated 14 months prior to the incidents and proper operation was.verified at that time. No corrective action or ~dministrative changes were cited by. the licensee. in' .his report.

( g) Safety Injection No. 10 occurr.ed wh i 1e the p1ant was in the process of coo 1down from Mode 3 * (hot standby) to Mode 5 (cold Shutdown).~j~h one reactor coolant pump operating. The steam generator (SG)

. 'atmospheric relief valves were being utilized to contro 1 the cool down, with the. operator monitoring

• the pressure in each of the four steam lines and

• adjusting the position of each SG atmospheric

• relief valve as necessary to maintain equal pres,.

sure in each steam line and obtain cooldown of the reactor coolant system.*

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With coolant_ pressure at 1500 psig and the reactor coolant Tave at 403°F, a steam line 6P trip occured causing a SI. Control room indication showed low pressure on one of the four steam lines with trips on two of the four steam line 6P chan-nels. A channel trip requires two out of three 6P

(>100 psig) signals between a steam line and the remaining three steam lines.

Since Salem, Unit 1 does not havP pressure re-cording instruments for the individual steam lines, the recorder traces of the reactor cool ant tempera-tures prior to and during the incident were evalu-ated, and do indicate that a 6P did exist between the steam generators of at least 60 to 70 psig. No instrument malfunctions were determined.

An evaluation of the incident indicates the fol-lowing contributing factors:

(1) "Use of the atmospheric relief valves for reactor cooldown instead of the steam dump system."

The use of the SG atmospheric relief valves to cool down the reactor is a complex and dif-ficult procedure. The operator is required to men itor each steam line pressure and i ndivi du-ally operate the four SG atmospheric relief valves to maintain equal pressure between SGs to obtain the desired cooldown rat?..

The Westinghouse-designed normal procedure for cooldown utilizes 12 steam dump valves ~on nec ted to a common steam header to a con-denser. A lead/lag controller (manual set 4

point) is used to opeh ~r modulate the required valves to regulate the steam dump rat~ and thus the reactor coolant cooldown rate. A swing check valve on each SG to .

header con- nee.ti on prevents backflow from the header and_ the other three SGs causing the heat* tr an sf er from the SGs to* be self~regulating. *Equal pressures will thereby be maintained in the SGs _and steam lines.

(2) "Use of only one reactor coolant pump when

• utilizing the SG atmospheric relief valves

- during cool down."

. Two pairs of inlet nozzles are located cin the opposite sides of the_ reactor vessel with. each noz~le of a pair having 45° of azimuthal separation. The four nozzles are not equally spaced, and with only one reactor coolant pump (RCP) in operation, backflow occurs in ttie - *

.other three cooling loops'and is the greatest in the nearest inlet nozzle. These different fl ow rates :make the task of contro 11 i ng the cooling rate of each SG difficult. By steaming through the atmospheric relief valve$, a subcooling effect is produced in the

_two lowest reverse flow loops causing. pressure i~ these loops to be lower. A slight adjust~

• -ment in the set po*; nt on any of the atmos-pheric reliefs under these conditions can easily cause a 100 psig ~P between loops re-sulting in a SI.

To prevent reotcurrences of th is last type of SI,*

• the licensee has proposed revising their pl ant Operating Instruction I-3.6, Hot Standby to Cold Shutdown,.*to reflect the following changes:

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1--

(l) Abnospheric relief valves will only be used for plant cooldown when the plant conditions prohibit use of the steam dump system (2) Specific direction will be provided in the procedure for monitoring steam generator pres-sur.es when use of the atmospheric reliefs is required

( 3)

• The procedure will require a minimum of two reactor cool ant pumps, and that they be di ag-onally opposed, to be in service during pl ant coo 1down.

The above changes will r'r' i-, effect prior to the next react or coo 1down.

(h) Injection 11 was caused when the output transformer and regulating resistors of Bus 18 inverter failed initiating a reactor trip. While recovering from the trip, a high steam flow-low* Tave safety in-jection occurred as Tave decreased below 543°F. As a result of this incident, a mis-aligned overspeed trip reset 1atch mechanism on an auxiliary feed pump and a faulty breaker for No. 11 RHR pump were found, corrected, and replaced, re-spectively. As of this date a formal ECCS Actu-ation Report has not been f i1 ed citing a cause, a corrective action, or precautionary measures to be taken to prevent an*injectionof this type from recurring.

3.0 CONCLUSION

3.1. Salem Safety Injections. Based on EI&C reviews of the Salem pl~nt LERs and ECCS Actuation Reports there is no common cause for the eleven Sis at Salem.

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The licensee-proposed procedural changes in regards to the use of atmospheric relief valves are satisfactory and will help prevent inad-vertent Sis simi 1ar to No. 10. It is further reconmended to the staff that pressure-recording instruments be added for each SG. Such instru-mentation will enable more accurate analysis of any-future Sis when the atmospheric relief valves must be used.

It is also concluded that the steps taken by the licensee to help prevent occurances similar to the first four are adequate.

Due to the abnormal number of personnel-error-caused Sis in this facility as compared to all other PWRs, a review of administrative controls and training procedures is recommended.

3.2 Generic Sis for PWRs. A review of the dockets for other PWRs for the past two years identified 12 other incidents of inadvertent Sis. These were distributed among several plants and resulted from an assortment of personnel errors and equipment malfunctions. The number and types of failures reported gave no indication that a generic problem of inadvertent Sis in PWRs exis.ts.

4.0 REFERENCES

]. PSE&G letter (Schneider) to NRC (O'Reilly) dated February 24, 1977.

(ECCS Actuation Report No. ECCS/77-01.)

2. PSE&G letter (Schneider) to NRC (O'Reilly) dated May 10, 1977.

(ECCS Actuation Report No. 77-26/990.)

3. PSE&G letter (Librizzi) to NR C (O' Re i 11 y ) dated .June 2, 1977.

(ECCS Actuation Report No. 77-29/990.)

4. PSE&G letter (Librizzi) to NRC (Grier) dated February 10, 1978.

(ECCS Actuation Report No. 78-04/990.)

5. Salem Unit 1, Technical Specifications, Appendix A to license DPR-70, August 13, 1976.

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