ML19296B859
| ML19296B859 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 02/11/1980 |
| From: | Wilson B Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19296B831 | List: |
| References | |
| NUDOCS 8002220271 | |
| Download: ML19296B859 (47) | |
Text
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UflITED STATES OF AMERICA flVCLEAR REGULATORY COMMISSI0fl BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
SACRAMENTO MUNICIPAL UTILITY DISTRICT
)
Docket No. 50-312 (SP)
(Rancho Seco Nuclear Generating Station)
)
NRC STAFF TESTIMONY OF BRUCE A. WILSON Ofl OPERATOR TRAINING AND COMPETENCE (Board - CEC Question 1-7, CEC Issue 3-1, CEC Issue 3-2, CEC Icsue 3-3, Board Question 32, and F0E Contention III(e))
Q.
Please state your name and N ur position with the NRC.
A.
My name is Bruce A. Wilson.
I am an employee of the U.S. Nuclear Regulatory Commission assigned to the Operator Licensing Branch.
Fr';. May 1979 until December 1979 I was with the Systems Group of the Bulletins and Orders Task Force.
Q.
Have you prepared a statement of professional qualifications?
A.
Yes. A copy of this statement is attached to this testimony.
Q.
Please state the nature of the responsibilities that you have had with respect to the Rancho Seco Nuclear Generating Station.
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i
, A.
I was responsible for reviewing part of SMUD's responses to the Commission Order of May 7, 1979. Specifically, I reviewed their procedures to ensure that their revised procedures were in accordance with t % requirements of the Order and complied with the Small Break Loss-of-Coolant Accident Guide-lines that were developed by B&W.
I also conducted an audit of some of Rancho Seco's operators and senior operators to evaluate the training they had received concerning the TMI-2 accident and the resulting impact at Rancho Seco.
Q.
What issues are addressed in this portion of your testimony?
A.
I am addressing Board - CEC Question.1-7 and CEC Issue 3-1, which state:
Board Question CEC-1-7 Do the operator training actions responding to Subpara-graph (d) of Subparagraphs a-e for Rancho Seco fail to give sufficient attention to providing appropriate analytical bases for operator actions?
CEC Issue 3-1 Whether personnel adequately understand the mechanics of the facility, basic reactor physics, and other funda-mental aspects of its operation?
Q.
Prior to the TMI-2 accident of March 28, 1979 what type of training did Rancho Seco licensed operators receive to assure their understanding of the mechanics of the facility, basic reactor physics, and other fundamental aspects of 1*., operation?
. A.
The procedures and criteria for issuing licenses to operators and senior operators are set forth by Commission regulations; 10 C.F.R. Part 55.
~
t1UREG-0094, "f1RC Operator Licensing Guide," is a guide that expands and explain; the regulations for obtaining a license. The specifics of the training program established by the Licensee to prepare candidates are contained in the Final Safety Analysis Report (FSAR); Section 12.3. This program was for the initial plant staff or " cold" license applicants. After the plant achieved criticality, the initial " hot" license applicants received the same training, while replacement operators recieved the training that is specified in Rancho Seco Topical Report T1-76, " Operator Training Program for Hot License Candidates."
In order to maintain a license all personnel must participate in the requalificati n program that is outlined in Rancho
$eco-procedure AP-25, " Licensed f4RC Operator Retraining." The cold, hot, and requalification training programs were reviewed and approved by the flRC.
The training of the Rancho Seco licensed personnal began in 1966 and continued through the licensing of the initial group of operators in 1974. More than one-half of the presently licensed personnel received all or most of the follow-ing training; several months observation at an operating nuclear plant, a twenty week course in basic reactor physics and engineering, a two month course in PWc achnology taught by B&W in Lynchburg, Va., and a six week simulatet u,rce d so taught by B&W.
In addition, these personnel participated in the staito activities of the unit which included testing components and systems and writing routine and emergency procedures.
f a The replacement operators participated in the hot license training program, which contains all the essentials of the cold program (described above) with several exceptions. Since the plant was operational, they were able to gain a great deal more practical training and therefore the observation training at another plant was deleted and the simulator course was shoretened.
Since December 1973 the Comission has required SMUD (and all other utilities) to have in effect a Requalification Program in which each licensed person must successfully participate in order to obtain a renewal of his license. The key aspects of the Rancho Seco program are the following: an annual written examination of comparable scope to the NRC test, an oral exam administered by facility management, a lecture series, assigned inidividual study, and a one week simulator course. Although attendance at the simulator cours' is not required by the Requalification Program, it has been SMUD's practice to send nearly all of their personnel every year. The few exceptions have bean meu ers of the management staff whose duties sometimes conflict with the simulator training. The Requalification Program is regularly audited by the NRC's Office of Inspection and Enforcement (I&E) and Operator Licensing Branch (0LB).
In the future, the requalification exams will be administered by OLB.
Q.
What additional training has been provided to Rancho Seco licensed operators pursuant to Subparagraph (d) of the short-term actions required by the Com-mission's May 7, 1979 Order?
. A.
To ensure that post-TMI information was adequately understood by Rancho Seco licensed operators, the following training and evaluations were performed:
1.
Each licensee has completed the TMI-2 sequence training on the simulator.
2.
Each licensee has successfully passed a SMUD administered.TMI related written examination, in which 90% was the passing grade.
3.
The above exams were audited for content and grading by the NRC.
4.
SMUD conducted special training sessions on the concepts and use of the small break LOCA procedure.
5.
Seven of the fourteen licensed personnel assigned to shift duty were audited by NRC.
6.
Several deficiencies revealed by the audit resulted in SMUD contracting with General Physics Corp. for additional training.
7.
An additional audit was conducted by General Physics (not by the individual who had administered the training).
8.
A followup audit of 8 operators was conducted by an NRC insputor, with no deficiencies uncovered.
. Q.
What steps has the t1RC taken to determine the Rancho Seco operators' level of understanding of the tra'
- g.
A.
Initial interveiws of Rancho Seco licensed personnel were conducted on June 1, 1979 (3 licensed personnel) and on June 2, 1979 (4 licensed personnel).
These interviews were conducted by myself and Philip Johnson, an inspector from_I&E Region V.
Q.
Did your interviews explore the operatord understanding of the analytical bases of actions which they may be required to take?
A.
Yes. The subjects covered were: TMI-2 Sequence of events, small break LOCA phenomenon, and the bases for changes to the licensee's LOCA procedures and other design and procedure changes made at Rancho Seco as a result of the TMI-2 accident. As a reference to discuss the analytical bases for the actions required in the small break procedure, Mr. Johnson and I used B&W's "Part II; Smal.1 Break Phenomenon - Description of Plant Behavior," a copy of which is attached hereto.
In particular, we used Figures 1 through 5 of the above document to determine if the licensed personnel were aware of the behav;or of the plant as a function of break size and equipment availability.
Q.
What were the test results, particularly on those portions related to operator's analytical understanding.
. A.
We found that the operators could satisfactorily explain the analytical basis for the small break phenomenon. We found, however, that there were some deficiencies in their knowledge of thermodynamics, natural circulation, and the TMI-2 sequence. These deficiencies could partly be attributed to the fact that some of the operators we interviewed had not yet attended the TMI-2 training session at the simulator.
In view of these deficiencies, the Licensee contracted with General Physics Corp,of Columbia, Md. to conduct additional training in these areas. This. training was audited separately by another employee of General Physics and re-audited by Mr. Johnson, who found no deficiencies in the analytical understanding of these phenomena among the eight licensed operators he audited.
Q.
On the basis of the ',ests that the NRC has conducted, do you believe that Rancho Seco licensed operators adequately understand the mechanics of the facility, basic reactor physics, and other fundamental aspects of its operation?
A.
Yes.
I conclude that Ranciso Seco operators adequately understand the mechanics of the facility, basic reactor physics, and other fundamental aspects of its operation.
Q.
On the basis of the tests the NRC has conducted, do you believe the Rancho Seco licensed operators adequately understand the analytical bases of the actions they may be required to take pursuant to Subparagraph (d) of the Commission's short-term required actions?
A.
Yes.
. Q.
What issue are you addressing in this portion of your testimony?
A.
I am addressing CEC Issue 3-2 which states:
CEC Issue 3-2 Whether personnel are properly apprised of new infor-mation pertinent to the facility's safe operation and ability to respond to transients, particularly infor-mation on operating experience of other reactors?
Q.
Does the licensee, SMUD, have a program for apprising its personnel of new information pertinent to the facility's safe operation and ability to respond to transients, particularly information on operating experience of other reactors?
A.
Yes. The licensee has stated that through the Requalification lecture series significant operating events at Rancho Seco and other facilities may be dis-cussed. Additionally, " Standing Orders," which shift supervisors are directed to discuss with their shift crews, may contain such information. Finally, when the licensed personnel participate in the annual simulator course at B&W in Lynchburg, Va. they are often exposed to events that have occurred on other B&W plants. See " Licensee's Answers (Set No. 2) To the California Energy Commission's First Set of Interrogatories Dated November 15, 1979," Answer to Interrogatory 22 (December 4, 1979).
. Q.
Does the NRC have a program for disseminating to reactor licensees, permittees, and applicants operational information from other licensed reactors?
A.
Yes. The NRC's Office of Management and Program Analysis (OMPA) has several means for dissiminating operational information. The first is a Licensee Event Report (LER) monthly listing. This is a computerized listing of LER's at each operating plant. Each LER is catorgorized as to cause (mechanical failure, human error, etc.) and there is a brief description of the event.
Secondly, OMPA publishes a document called " Power Reactor Events" in which signficant events which could have generic implications are described.
Upon a licensee's request, it can receive copies of these documents.
Special printouts of LER's may also b_e requested by the individual licensees.
OMPA also distributes the Gray Book, " Operating Units Status Report," which is sent to all licensees that have submitted input for it.
Q.
Has the NRC undertaken any efforts to improve the dissemination of operational information?
A.
The Commission has established an agency-wide Operational Data Analysis and Evaluation Office to provide coordination and an overview of all operational data analysis - related activities performed within the NRC. The individual program offices have also been directed to establish operational data analysis capability.
. Q.
Has the nuclear industry undertaken a program for the review of plant event reports and data?
A.
Yes. The Electric Power Research Institute (EPRI) has founded a Nuclear Safety Analysis Center to systematically review available plant event reports and data. Also, the industry has established the Institute for Nuclear Power Operations (INPO). One of the functions of INPO is to review and analyze nuclear power plant operating experience and feed this information back to the utilities. The utilities can then incorporate this information into the training programs.
Q.
Ic the NRC considering further iequiretents for imposition on licennees regard-ing dissemination of operating experience to their personnel.
A.
Yes. The Commission is considering imposition of a requirement that licensees review their administrative procedures to assure that operating experience from within and outside their organizations is continually provided to operators and other operations personnel and is incorporated into training programs. Draft NUREG-0660, Action Plans For Implementing Recomendations of the President's Comission and Other Studies of TMI-2 Accident (12/10/79), Task I.E.2.
Opera-ting plant licensees would be required to have completed this task by September 1980.
. Q.
Based on the above programs, do you believe SMUD's personnel are now being properly apprised of pertinent new information?
A.
I believe the Licensee has a program through which its personnel can be apprised of pertinent new information. Additional requirements may be imposed by the NRC on licensees with regard to dissemination of operating experience. The NRC Staff believes that substantial improvement can be made in the process of dissemination of operating experience. However, based on my audits of licensed personnel at Rancho Seco, I conclude that they have an adequate understanding of the implications of the TMI-2 accident. The licensee's program of disseminating information on the TMI-2 accident has, I therefore conclude, been successful in enabling its operators to understand the implications of th0t accident.
Q.
What issue is addressed in this portion of your testimony?
A.
I 6m addressing CEC Issue 3-3, which states CEC Issua 3-3 W. ether NRC and SMUD adequately ensure that emergency instructions are understood by and are available to plant personnel in a manner that allows quick and effective implementation during an emergency?
Q.
Please describe the organization of the Licensee's Emergency Procedures.
A.
The Licensee's Emergency Procedures (EP's) are generally divided into six sections: Purpose, Description, Symptoms, Automatic Actions, Immediate
. Operator Actions, and Subsequent Operator Actions. During an emergency situation, the licensed operators must diagnose the event by matching the plant parameters with the Symptoms as listed in the EP's.
They must then ensure that the Automatic Actions have occurred and take the required Immediate Operator Actions. These three steps must be done by memory.
The operator should get out the appropriate procedure, ensure that the above three steps have been accomplished correctly and then follow the instructions listed under Subsequent Actions.
Q.
Are the EPs available in a manner that allows quick and effective implementa-tion during an emergency?
A.
Yes. The Licensee's emergency procedures are contained in a red book in a desk drawer immediately behind the control console in the control room.
Q.
Does the Licensee have procedures to ensure that procedures are kept up-to-date?
A.
Yes. Administrative procedures exist that are intended to ensure that these procedures are kept up-to-date. The Requalification Program also covers the latest procedure revisions.
. Q.
How does the NRC determine whether licensed personnel have an adequate understanding of EPs?
A.
Through the examination process, the NRC determines whether EP's are under-stood by licensed personnel. Applicants are asked on the written examination to write down those portions of selected emergency procedures that must be committed to memory. On the oral examination, the applicants are asked to simulate or " walk through" these procedures and demonstrate to the examiner their familiarity with and understanding of these procedures.
Questions concerning every EP are not asked of each applicant.
It is an audi't process, as is the remainder of the oral and written examination.
Typically, two of the EP's will be on the written examination, three or four will be discussed in the control room during the oral examination and several more during the walk-through in the plant. The examiner will cover different EP's in the oral examination of other applicants.
In this way, the examiner can covdr all or most of the EP's.
The knowledge and use of emergency procedures is always included as a topic on the exit interview that is conducted between the examiner (s) and the licensee's management. On the basis of the examinations conducted to date at Rancho Seco, the NRC is satisfied that licensed personnel understand the emergency procedures.
. Q.
Does the Licensee have a program for determining that licensed personnel have a continuing understanding of EPs?
A.
Yes. Through the Requalification Program, the licensed personnel must demonstrate continuing understanding of EPs. Section 3.2.1 of the Requalification Program requires the following:
... each licensed Senior Operator or Operator shall participate in an oral examination with the plant superintendent or his designated representative. This examination and evaluation shall contain the following:
1.
A discussion of required actions durino abnormal or emergency conditions.
2.
A simulation of abnormal and emergency conditions while in the Control Room showing each action and controlling device to be operated.
3.
Should the performance of the licensed Senior Operator or Operator be deemed unsatisfactory, the Senior Operator or Operator will partici-pate in an accelerated review program tailored to place emphasis where there is clear indication of need.
4.
Upon completion of ',he accelerated review program, the individual shall be subject '..a re-examination.
SMUD has made this oral e.iination an annual requirement. This exceeds the requirements of Appendix A, 10 C.F.R. 55.
. Q.
How were Rancho Seco emergency procedures changed as a result of TMI-2 and the M;y 7, 1979 Commission Order.
A.
The Comission Order required all B&W licensees to develop and implement operating instructions to define operator actions for potential small break loss-of-coolant accidents (SBLOCA). B&W then developed guidelines to be used in the rewrite of the LOCA emergency procedures. With the use of these guidelines, the Licensee rewrote EP D.5 " Loss of Reactor Coolant /
Reactor Coolant Pressure." The NRC staff n riewed this revised procedure to ensure that it conformed to the guidelines. We also " walked-through" the procedure in the Rancho Seco control room to ensure that the steps were in a logical order and that the instruments and controls were readily avail-able for the operators to perform the required tasks. On the basis of this review we were satisfied that the revised procedure met the requirements of the Commission Order.
Q.
Were any other emergency procedures changed?
A.
Yes. Nearly all of the emergency procedures have had some revisions in the last few months. Most notably, EP D.14 " Loss of Steam Generator Feed" was revised to include actions to be taken in the event all feedwater was lost for an t. tended period of time. Thi procedure, and several others, incorporate the 50 F subcooling criteria. Emergency Procedure D.1, " Load Rejection," D.2,
" Turbine Trip," and D.3, " Reactor Trip" were revised to include the new turbine trip - reactor trip circuitry. Finally, all of the EP's were revised to include a reminder to the operators to check alternate instrument channels of key parameters.
On the basis of our review, we believe the Licensee has made significant improve-ments to the emergency procedures.
. Q.
What contentions does this portion of your testimony address?
A.
I am addressing F0E Contention III(e), which states:
F0E Contention III(e)
The NRC orders in issue do not reasonably assure adeauate safety because no procedures exist or have been taken for the determination of the adequacy of operator competence.
I am also addressing Board Question 32 insofar as it relates to the competence of licensed personnel at Rancho Seco. Board Question 32 states:
Board Ques + ion 32 Rancho Seco, being a Babcock and Wilcox designed reactor, is operated by personnel and management whose-competence has not been adequately tested and evaluated, namely testing has not been con-ducted as to whether such employees can act responsibly and appropriately to make judgment decisions during a loss of feedwater transient, personnel interviews have not been conducted to properly evaluate the test results with such employees and some employees have never been tested because of grandfathering, and therefore is unsafe and endangers the health and safety cf Petitioners, constituents of Petitioners and the public.
Q.
Does the Licensee, SMUD, have a program and procedures for testing the competence of its operating personnel?
A.
The response to CEC 1-7 and CEC 3-1 contains information regarding the initial training, retraining and evaluation of Rancho Seco licensed personnel.
Included
. was an outline of the Licensee's approved Requalification Program in which the operators receive an annual facility administered written and oral examination. These programs satisfy the present NRC requirements for test-ing the competency of operating personnel.
Q.
In the period since the TMI-2 accident on March 28, 1979 what steps has the Licensee taken to test and evaluate the comoetence of its operating personnel to act responsibly and appropriately to make judgment decisions during a loss of feedwater transient?
A.
The respcnse to CEC 1-7 and CEC 3-1 listed the additional training and evalua-tions conducted since March 28, 1979. This training included diagnosing and responding to a loss of main feedwater transient.
In addition, emergency procedure 0.14 " Loss of Steam Generator Feed" was revised to contain guidance for the operators to respond to a complete loss of feedwater. By virtue of their participation in simulator training, all of the licensed operators have received additional trairiing and lectures on loss-of-feedwater transients.
Q.
Does the Licensee conduct interviews with its operating personnel to discuss their performance on the tests administered?
A.
Yes. The Requalification Program has provisions for discussing test perfor-mances with the licensed operating personnel.
. Q.
Has the Licensee exempted any of its licensed personnel from being tested to determine their ability to make proper judgment decisions during a loss of feedwater transient (i.e., " grandfathered" those personnel)?
A.
Literally, the answer to this question is yes, in that the licensee's training coordinator, who wrote and graded the licensee administered examina-tions concerning TMI-2, was not required to take that examination.
Pursuant to the licensee's approved Requalification Program, the person writing the annual written requalification examination plus a maximum of two others, who may assist in its preparation and grading, are exempt from that examination.
Q.
Briefly summarize the history and results of licensed operator testing by the NRC at the Rancho Seco facility.
A.
Since May 1974, a total of twenty six applicants have been examined and subseq-uently licensed at Rancho Seco. Eighteen originally applied for a complete senior operator examination, i.e. operator written, scnior operator written, and an oral examination. All of the eighteen passed the examinations on their initial attempt. Eight other applicants have applied for reactor operator licenses since January 1975. Two of these applicants failed initially, but passed a subsequent examination within one year and were issued reactor operator licenses.
Four of these eight licensed operators subsequently applied for upgrade to senior operator. One of these applicants failed the initial examination, but passed a subsequent one.
. Q.
As a result of the Commission's May 7, 1979 Order, what additional testing of licensed personnel at Rancho Seco has been conducted by the NRC?
A.
The additional testing of licensed personnel at Rancho Seco that has been conducted by the NRC includes the following:
1.
Oral interviews by an OLB examiner and an I&E inspector of 7 licensed personnel on June 1 and 2, 1979.
2.
Re-audit of 8 licensed personnel by the same I&E inspector on June 7 and 8, 1979.
3.
Written and oral senior operator examinations administered to a licensed Rancho Seco operator by an OLB consultant examiner on November 29, 1979.
Q.
Would you say that these NRC tests have covered whether the licensed operators can act responsible and appropriately to make judgment decisions during a loss of feedwater transient?
A.
Yes. The attached letter from J. J. Mattimoe, SMUD, to Harold R. Denton, NRC was used as an aide by NRC personnel conducting the audits in June, 1979. The following subjects were covered in the control room with the Rancho Seco licensed operators:
1.
Verifying AFW flow on loss of 4 RCF's (pg. 1).
2.
How to power AFW pumps from essential Nuclear Services buses 4 A/4B (pg. 1).
. 3.
Reason for stationing an operator at FW Valve-055 during surveillance tests (pg. 2).
4.
AFW values that have been added to the locked valve list (pg. 2).
5.
Control of AFW ficw independent of ICS (pg. 3).
6.
Changes to emergency procedure D.14, " Loss of Steam Generator Feed" (pg. 3).
7.
Modifications to AFW flow indications (pg. 4).
8.
Procedure for transferring AFW pump suction to alternate supply (pg. 5).
9.
Changes to emergency procedure D.10, " Loss of Reactor Conlant Flow /RCP Trip" (pg. 5).
- 10. Changes to control room annunciators for all auto start conditions of the AFW system (pg. 5).
Also attached to Mr. Mattimoe's letter is the lesson plan for instruction of licensed personnel.
During the audits conducted in June 1979 by the NRC, no deficiencies were found in the licensed operators' ability to respond responsibly and appropriately to a loss of feedwater transient.
Q.
Did the interviews conducted on June 1-2, 1979 reveal any other areas of weakness?
A.
Yes. Certain operators displayed insufficient comprehension of thermodynamics, natural circulation, and the TMI-2 sequence.
Q.
Has the NRC conducted follow-up interviews with these individual licensed personnel to discuss these areas of weakness?
A.
Yes. On June 17 and 18, 1979 Mr. Philip Johnson of Region V conducted eight follow-up interviews. The follow-up interviews demonstrated substantially improved knowledge in the:e areas. On the basis of the follow-up interviews,
~
Mr. Johnson found tlie operators' comprehension in these areas to be adequate.
Q.
On the basis of your review of the Licensee's training and testing program, do you believe the Licensee has effective procedures for determining the competence of its operating personnel?
A.
Yes.
I believe the licensee's present procedures are effective for deter-mining the competence of the operating personnel.
BRUCE A. WILSON PROFESSIONAL QUALIFICATIONS I am a Reactor Engineer in the Operator Licensing Branch, Division of Project Management, Office of Nuclear Reactor Regulation.
I am responsible for developing,-
preparing and administering examinations for applicants for reactor operator and senior reactor operator licenses.
I.am assigned to the Power and Research Reactor Group, which is primarily responsible for administering examinations on Combustion Engineering and Babcock & Wilcox designed reactors in addition to research reactors.
I received a Bachelor of Science Degree in Mechanical Engineering in 1966 from Syracuse University and a Master of Science in Nuclear Engineering in 1967 from the University of Washington.
In 1967, I entered active duty with the United States Air Force and was assigned to the 10 Megawatt Nuclear Engineering Test Facility (NETF), Wright Patterson AFB, Dayton, Ohio. From 1967 to 1968. I was a ~ Project Engineer in the Experimental Branch where my primary function was to design and perform safety analyses of in-core irradiation test experiments.
From 1968 to early 1970 I was Chief, Reactor Engineering Section, where I performed safety analyses for reactor modifications and safety limit bases for technical specifications. During this period, I was certified as a Reactor Operator and Shift Supervisor at the NETF by the Air Force Directorate of Nuclear Safety.
From 1970 to 1971 I was assistant to the Chief, Operations and Maintenance Division during the final decommissioning and entombment of the facility.
In 1971, I was transferred to the Armed Forces Radiological Pesearch Institute in Bethesda, Maryland.
For eight months, I was Project Manager in the Accelerator Division and then transferred to the Reactor Division, where I was Assistant Physicitt-in-Charge of a TRIGA Mark F reactor.
I received a Senior Reactor Operator's License for this facility from the U.S. Atomic Energy Comission (AEC) and was primarily responsible for experiment safety review, technical specification revision and training.
In October 1973, I resigned my commission with the Air Force and joined the Operator Licensing Branch of the AEC. From May to December 1979, I was assigned to the Systems Group of the Bulletins & Orders Task Force.
My functions on this Task Force were to review and approve the Small Break Loss-of-Coolant Accident (SBLOCA) Guidelines developed by Westinghouse and B&W, and to insure that the applicable facilities have developed emergency pro-cedures incorporating these Guidelines. Finally, I audited the operators and training records to determine that sufficient training had been conducted regarding the SBLOCA phenomenon and the revised emergency procedures.
r Part II: Small Break phenomena - Description of Plant Behavior 1,
Introduction A loss-of-coolant accident is a condition in which liquid inventory is lost from the reactor coolant system. Due to the loss of mass from the reactor coolant system, the most significant short-term symptom of a loss-of-coolant accident is an uncontrolled reduction in the reactor coolant system pressure The reactor protection system is designed to trip the reactor on low pressure.
This should occur before the reactor coolant system reaches saturation conditions. The existence of saturated conditions within the reactor system is the principal longer-term indication of a LOCA and requires special consideration in the development of operating procedures.
Following a reactor trip, it is necessary to remove "ecay heet from the reactor core to prevent damage. However, so long as the reactor core is kept covered with cooling water, core damage will be avoided. The ECCS systems are designed to respond automatically to low reactor coolant pressure conditions and take the inital actions to protect the reactor core. They are sized'to provide sufficient water to keep the reactor core covered even with a single failure in the ECCS systems. Subsequent opera 6or actions are required ultimately to place the plant in a long-term cooling mode. Thecherall
. objective of the automatic emergency core cooling system and the followup operator actions is to keep the reactor core cool.
A detailed discussion of the small break LOCA phenomenalogy is presented in this section. This discussion represents Part II of the operating procedure guidelines for the development of detailed operating procedures. Part I presents the more detailed step-by-step guidelines.
c,
The response of the primary system to a small break will greatly depend on break size, its location in the. system, operation of the,
reactor coolant pumps, the number of ECCS trains functioning, and the ava.11 ability of secondary side cooling. RCS pressure and pressurizer level histories for various combinations of parameters are presented in order to indicate the wide range of system behavior which can occur l
for small LOCA's.
~
2.
Impact of RC pump Ooeration on a Small LOCA
'With the RC pumps operating during a small break, the s' team and water will remain mixed during the transient. This will result in liquid being discharged out the break continuously. Thus, the fluid in the RCS can evolve to a high void fraction as shown in Figure 1.
, The maximum void fraction that the system evolves to, and the time it occurs, is dependent on the break size and location. Continued RC pump operation, even at high system void fractions, will provide sufficient core flow to keep cladding temperatures within a few degrees of the saturated fluid temperature.
Since the RCS can evolve to a high void fraction for certain small oreaks with the RC pumps on, a RC pump trip by any means (i.e., loss of offsite power, equipment. failure, etc.) at a high void fraction during the small break transient may lead to inadequate core cooling. That is, if the RC pumps trip at a time period when the system void. fraction is greater than approximately 707,, a core heatup will occur because the am.ount of water left in the RCS would not be sufficient.to. keep the c6re covered. The cladding temperature would increase until core cooling is re-established by the ECC systems. For certain break sizes and times of RC pump trip, aceeptable peak cladding temperatures during the event could not be assured and the core could be damaged..Thus, prompt operator action to trip the RC pumps upon receipt of a low pressure ESFAS signal is required in order e
3 to ensure that adeqJate core cooling is provided. Following the RC pump trip, the small break transient will evolve as described in the subsequent sections.
3.
Small Breaks with Auxiliary Feedwater There are four basic classes of break response for small breaks with auxiliary feedwater. These are:
,1.
LOCA large enough to depressurize the reactor coolant system 2.
LOCA which stabilizes at approximately secondary side-pressure 3.
LOCA which may repressurize in a saturated condition 4.
Small LOCA which stabilizes at a primary system greater than secondary system pressure The system transients for these breaks are depicted in Figure 2.
3.1 LOCA large Enouah to Deoressurize Reactor Coolant System: ' Curves
~
1 and 2 of Figure 2 show the response of RCS pressure to breaks that are large anough in combination with the ECCS to depressurize the system to a stable low pressure. ECCS injection easily exceeds core boil-off and ensures core cooling. Curves 1 and 2 of Figure 3 show the pressurizer level transient. Rapidly falling pressure causes the hot legs to saturate quickly. Cold leg temperature reaches saturation somewhat later as RC pumps coast down or the RCS depres'surizes below the secondary side saturation pressure. Since these breaks are capable of depressurizing the RCS without aid of the steam generator.s,.they are essentially unaffected 4y the a.vailability of auxiliary feedwater. Upon receipt of a low pressure ESFAS signal, the operator.must trip all RC pumps and perify that all ESFAS attens have been completed. The operator must also balance HPI flows such
~
that flow is available through all HPI, injection nozzles even if only one HPI is available.. The operator should also balance LPI flows, should the system be actuated, to ensure-flow through both. lines. The. operator -
needs to take no further; actions _to bring the system to a.saf.e shutdown 3
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condition. Rapid depressurization of the steam generators would only act to accelerate RCS depressurization. It is, however, not necessary.
Restarting of the RC pumps is not desirable for this class of break.
l.ong-term cooling will require the operator to shift tne LPI pump suction to the reactor building sump.
3.2 LOCA Which Stabilizes at Approximately Secondary Side Pressure. Curve 3 of Figure 2 shows the pressure transient for a break which is.too small in combination with, the operating HPI to depressurize the RCS. The steam generators are, therefore, necessary to remove a portion of c. ore decay heat. Although the system pressure will initially stabilize near the secondary side pressure, RCS pressure may eventually begin falling as the decay heat. level decreases. Curve 3 of Figure 3 shows pressurizer level b*ehavior. The hot leg temperatur'e quickly equalizes to the saturated temperature of the secondary side and controls primaiy system pressure at saturation. The colt leg temperature may remain slightly subcooled.
If the HPI refills and repressurizes the RCS,-the hot legs can become subcooled. The immediate operator action is to trip the RC pumps upon receipt of the low pressure ESFAS signal' and then verify ESFAS functions.
The operator must then balance HPI in order to ensure flow through each high pressure injection.line.
Followup action by the operator is to raise the emergency feedwater level to 95% on the, operating range and check for established natura~l circulation. This is done by gradually depressurizing the steam generators. If this test " fails, intermittent bumping of a RC pump should 6e perfomed as soon as one N available.. Continued depressurizatior, of the steam generators with natural circulation leads to cooling and
- 3
~depressurization of the RCS. The operator's goal is to depressurize. the RCS~to a pressure that enables the ECCS to exceed core boil-off, possibly refill the RCS, and tp ultimately establish long-term cooling.
a.
e
3.3 LOCA Which May Repressurize in a Saturated Condition.
Curve 4 of Figure 2 shows the behavior df a small break that is too small, in combination with the HpI, to depressurize the primary system. Although steam generator feedwater is available, the loss of primary system coolant and the resultant RCS voiding will eventually lead to interruption of natural circulation. This is followed by gradual repressurization of the primary system.
It is possible that the primary system could repressurize as high as the pressurizer safety valve setpoint before the pressure stabilizes.
This is shown by the dashed line in Curve 4.
Once enough inventory has been lost from the primary system to allow direct steam condensation in the regions of the steam generators contacting secondary side coolant, the primary system is forced to depressurize to the saturation pressure of the secondary side.
Since the cooling capabilities of the secondary side are neeced to continue to remove decay heat, RCS pressure will not fall below that on the secondary side. HPI flow is sufficient to replace the inventory lost to boiling in the core, and condensation in the steam generators removes decay. heat energy. The RCS is in a stable thermal conditici and it will remain there until the operator takes further action. The pressurizer level response is characterized by Curve 3 of Figure 3 during the depressurization, and Curve 4 of Figure 3 during the temporary repressurization phase. The dashed line indicates 'the level behavior if pressure is forced up to-the pressurizer safety valve setpoint. During this transient, hot leg temperature will rapidly approach saturation with 'thel initial system depre'ssurization, and it will. remain saturated during the whole transient.
Cold leg temperature will approach 'saturatifon as circulation 'is lost, but
~
niay remain 'slightly subco'oled 'during the'repressurization phase of the transient. L'ater RCS de~ press'urizatic'n'c'ould cause '1!he c'tild 1eg % '
~
~
~
temperatures to reac'h,ssturation. Subsequent refilling of the primary e
system by the HPI might cause temporary interruption of steam condensation in the steam generator as the primary side level rises above the secondary side level.
If the depressurization capability of the break and the HPI is insufficient to offset decay heat, the primary system will once more repressurize. This decreases HPI flow and increases loss through the break until enough RCS coolant is lost _to once more allow direct steam l
condensation in the steam generator. This cyclic behavior wi_11 stop
~
once the HPI and break can balance decay heat or the operator takes some action.
The operator's irreadiate action is to trip the RC pumps upon receipt of the low pressure ESFAS signal and verify the completion of all ESFAS functions. The operator should then balance HPI flow.
Following that, he should raise the steam generator level to 95% of the operating range and check for natural circulation.
If it is positive, he should depressurize the steam generators, cool and depressurize the primary system, and attempt 1
to refill it and establish long-term cooling.
If the system fails to go into natural circulation, he should open the PORY long enough to bring and hold the RCS near the secondary side pressure. Once natural circulation is established or a RC pump can be bumped, he will be able to continue depressurizing the RCS with the steam generators and establish long-term cooling.
3.4 Small LOCA Which Stabilizes at P> psec. Curve 5 of Figure 2 shows the behavior of the llCS ' pressure to a break for which high pressure injection is being supplied and exceeds the leak flow before th) pressurizer has emptied. The primary system remains subcooled and natural circulation to the steam generator removes core decay heat. The pressurizer never empties and continues to control primary system pressure. The operator needs to trip theRC'pumpsandensuretNatESFASac3ionshaveoccu[ed. Throttling of HPI is permitted only after RCS subcooliniof 50 F has been established, the 0
~
pressurizer has refilled, and natural or forced circulation has been
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A restart of the RC pumps under these conditi Ln p.; W]
for plant control.
ons is desirable 3.5 Small Breaks in Pressurizer."
break in the pressurizer will behave in a maThe sy
~
' c-A 3.N; h
[q discussed. The initiai depressurization, howevernner simila jd the initial inventory loss is entirely steam, will be more rapid as
. 3
.iJ'g3 The pressurizer level res'ponse for these accid
~1 like a very small break without auxiliary feed ents will initially behave water.
4 '!.? l pressurizer level shown in Figure 4 will occur d The initial rise in 4
2 (k.
reduction in the pressurizer and an insurge of ue to the pressure Lib from the RCS.
coolant into the pressurizer Once the reactor trips, system contraction ca
,.a level in the pressurizer.
- +
uses a decreasing
(
piping and cause an insurge into the pressur. F g
eg izer.
the pressurizer.
This ultimately fills ff For the remainder of the transient a
wi.1 remain full.
, the pressurizer Toward the later stages of the transient 3,
may contain a two-phase mixture and the indicated l
, the pressurizer 5
pressuri:... ~
evel will show that the
)(
s only parti. ally full. 'Except for closing the PORV operator actions and system response are the same f ock valve, l
similar breaks in the loops.
or these breaks as for l
[
4.
Small Breaks Without Auxiliary Feedwater i
There are three basic classes of break respons without auxiliary [feedwater. These are:
e for small breaks 1.
Those breaks capable of relieving all decay he t 2.
via the break.
a Breaks that relieve decay heat with both the HPI i
'njection and via the break.,
~
3 Breaks which do not automatically actuate the'HPI'
. =.
system repressurization.
and result in
. The system pressure transients for these bre' ks '
a are depicted in Figure S.
4
8 4.1 LOCA's Large Enough to Depressurize Reactor Coolent System. For Class 1 (curve 1 of Figure 5), RC system pressure decreases smoothly throu'hout the transient. For the larger breaks in this class, CFT g
actuation and LPI injection will probably occur. For the smaller breaks of this class only. CFT actuation will occur. Auxiliary feedwater injection is not necessary for the short-term stabilization of these breaks. The pressurizer level for this transient rapidly falls off scale. Operator action and plant response are similar to those described for this class of breaks with a feedwater supply.
4.2 LOCA's Which Reach a Semi-Stabilized State. For Class 2 (Curve 2 of Figure 5) breaks, the RC pressure will rapidly reach the low pressure ESFAS trip signal (about two to three minutes 1. With the HPI's on, a slow system depressurization will be established coincident with the decrease incoredecayheht. No CFT actuatien is ex'pected. Auxiliary feedwater is not necessary for the short-term stabilization of these breaks. The pressurizer level for this transient rapidly falls off scale.
The operator needs to trip the RC pumps upon the low pressure ESFAS signal, verify completion of all ESFAS functions, and try to establish secondary side c'coling. ' Balancing of the HPI must also be perfonned.
If steam generator feedwater cannot be obtained and RCS pressure is increasing, the operator should open the PCRV and provide all the HPI and makeup capability possible. The goal is to depressurize and cool the core with the ECCS, the PORV, and the break.
If secondary side cooling is again established, the operator should verify natural circulation, and if unavailable, bump a RC pump to complete RCS cooldown with the steam generators. At this point, the PORV can be closed, the system refilled, and long-term cooling established.
G
. me..
_g_
4.3 Small LOCA's Which do not Actuate the ESFAS. Automatic ESFAS actuation will not occur for Class 3 (Curve 3 of Figure 5) breaks.
Once the SG secondary side inventory is boiled off, system repressurization will occur as the break is not capable of removing all the decay heat being
~
generated in the core. Cystem repressurization to the PORV or the pres-surizer safety valves will occur for smaller breaks in this class.
For the "zero" break case, repr:assurization to the PORV will occur in the first five minutes. Operator action is required within the first 20 minutes to ensure core coverage throughout the transient. For the 177-FA lowered-1990 plants, this action can be either manual actuation of the auxiliary feedwater system or the HPI system.
The establishment of auxiliary feedwater will rapidly depressurize the RCS to the ESFAS actuation pressur6, and system pressure will stabilize at eiLher the secondary side SG pressure or at a pressure where the HPI equals the leak rate. Upon receipt of the low pressure ESFAS signal, the operator must trip the RC pumps.
For the raised loop Davis-Besse plant (which has a safety-grade auxiliary feedwater system) operator action is necessary at some time greater than 20 minutes (approximati:1y 40 minutes) as there is increased inventory in the loops that is available to drain into the reactor vessel.
Hcwever, because the plant is equipped with low shut-off head HPI pumps, the operator must establish auxiliary feedwater in order to depressurize the RCS.
For the Class 3 breaks, pressurizer level respons'e will be as shown in Figure 6.
The minimum refill time for the pressurizer is that for the "zero" break and is shown in Figure 6.
After initially drawing inventory from the pressurizer, the system repressurization will cause the pressurizer
~
level to increase, possibly to full pressurizer level. Once the operator action to restore auxiliary feedwater has been taken, the system a
depressurization will result and cause an outsurge from the pressurizer.
Complete loss of pressurizer level may result.
For the smaller breaks in Class 3 which result in a system repressurization following the actuation of the HPI system, pressurizer level will increase and then stabilize.
Without auxiliary feedwater, both the hot and cold leg temperatures will saturate early in the transient and, for the Class 1 and 2 breaks, will remain saturated. For the Class.3 breaks, once auxiliary feedwate 6
i is established, the cold leg temperatures will rapibly decrease to f
approximately the sautration temperacure corresponding to the SG secondary side pressure and will remain there throughout the remainder o'f the transient. Hot leg temperatures will remain saturated throughout the event.
The operator needs to manually initiate all ESFAS actions, balance HPI
~
flow, and attempt to restore secondary side cooling. ~ In the meantime, he a
should actuate the makeup pump and open the PORY in order to cool the core and limit the RCS repressurization. Once feedwater is available, he can close the PORV and continue the RCS cooldown 'and depressurization with the
)
If natural circulation has net been established, he ;
can bump a RC pump to cause forced circulation. The goal is to depressurize to where the ECCS can refill the RCS and guarantee long-term cooling.
4.4 Small Breaks in Pressurizer. See the writeup for. small breaks in pressurizer with feedwater.
Small breaks in the pressurizer will differ from those in the loops in the same manner as those previously described in the section addressing small breaks in the pressurizer with auxiliary feed..
5.
Transients with Initial Response Similar to a Small Break Several transients give initial alams similar to small breaks. These transients will be distinguished by additional alams and indications or subsequent system response.
Overcooling transi,ents such as steam line weaks, increased feedwater
-.e
flow, and steam generator overfill can cause RCS pressure decreases with low-pressure reactor trip and ESFAS actuation. But steam line breaks actuate low steam pressure alarms for the affected steam generator, and steam generator overfills result in high steam generator level indications.
The overcooling transients will repressurize the primary system because of HPI actuation, and will return to a subcooled condition during repres-surization. The immediate actions for both overcooling and small break
~
transients are the same, including tripping of the RC pumps.
The operator will recognize overcooling events during repressurization, if not sooner, and is instructed to throttle HPI and restart the RC pumps, if subcooled conditions are established, by the small brea.k operating instructions.
A loss-of-feedwater transient will result in a high reacter system pressure alarm but does not give an ESFAS actuation alarm.
A loss of integrated control system power transient starts with a hioh RC pressure trip. After the reactor trip, this becomes an overcooling transient and will give low reactor system pressure and possible ESFAS actuation. Steam generator levels remain high and the system becomes subcooled during repressurization.
Design features of the B&W NSS proside automatic protection during the early part of small break transients, thtreby providing adequate time for small breaks to be identified and arpropriate action taken to protect the system. The only prompt manual operator action required is to trip the RC pumps once the low pressure ESFAS signal is re} ached.
6.
Trans-ients that micht initiate a LOCA There are no anticipated transients that might initiate a LOCA since the PORY has been reset to a higher pressure and will not actuate during anticipated transients such as loss of main fecdwater, turbine trip, or loss of offsite power.
e
r 12-
- e However, if the PORV should lif t and fail to reset, there are a number of indications which differentiate this transient from the anticipated transients identified above. These include:
o ESFAS actuation o Quench tank pressure / temperature alams o Saturated primary system o Rising pressurizer level l
These additional signals will identify to the operator that in addition to the anticipated transient, a LOCA has occurred.
In the unlikely event that small' breaks other than a malfunctioning PORV. occur after a transient, they can be identified by initially decreasing RCS pressure i
andconvergencetosaturationconditionsinthereactorcoolant. Small break repressurization, if it occurs, will follow saturation conditigns.
By remaining aware of whether the reactor coolant remains subcooled or becomes saturated after transients, the operator is able to recognize when a small break has occurred.
7.
HPI Throttling For small LOCA's, the HPI system is needed to provide makeup to the
~
RCS and must remain operable unless specific criteria are satisfied.
The basis for these criteria are described below.
For certain small breaks, system depressurization will result in LPI actuation. Since the LPI is designed to provide infection at a greater' capacity than the HPI, termination of the HPI is allowed. However, this action should only be taken'if the flow rate through each line is ati least 1000 gpm and the situation has been stable for 20 minutes. "The 20-minute time delay is included to ensure that the system will 'not repressurize and result in a loss of the LPI fluid.
In the event of a core flooding line break, the LPI fluid entering the broken core flooding line will not reach the vessel. Thus, in order to ensure that fluid is continually being injected to the RV for all breaks, the LPI must be providing fluid through both lines. The 1000 gpm is equivalent to the flow from 4
- W.
two HPI pumps and ensures that upon termination of the HPI pumps, adequate flow is being delivered to the RV.
Throttling or termination of the HPI flow is also allowed if all the following criteria are met:
0 A.
Hot and cold leg temperatures are at least 50 F below the saturation temperatures for the existing RCS pressure.
0 B.
Hot leg temperatures are no more than 50 F hotter than thi secondary side saturation temperature (This ensures that heat is being removed via the SG.)
C.
The action is necessary to prevent the indicated pressurizer level from going off-scale high.
Under these conditions, the prima y system is solid. Continued HPI flow at full capacity may result in a solid pressurizer and would result in a lifting of the PORV and/or the pressurizer code safety valves.
This i
may in turn lead to a LOCA. Thus, HPI flow should be throttled to 0
maintain a stable inventory in the RCS. However, if the 50 F subcooling cannot be maintained..the HPI shall be immediately reactivated.
HPI flows should also be throttled to prevent violation of the nil ductility temperaturc (NDT) for the reactor vessel.
This concern can only arise if the fluid temperature 0
within the reactor ve,ssel is at least 50 F subcooled. A curve of the allowable downcomer temperature for a given RCS pressure is provided within the operating guidelknes. The downcomer temperature is determined by ope ~of two methods:
1.
If one or more RC pumps are operative, the cold leg RTD reading will be essentially the same as the reactor vessel downcomer temperature.
2.
Without the RC pumps operating, the cold leg RTO's may not provide e
14 temperature readings indicative of the actual RV downcomer temperature, as a stagnant pool of water may exist at these locations. The incore thermocouples will provide the best indicator of the downcomer temperature and should be utilized if no RC pumps are available.
In order to account for heat 0
added to the fluid from the core,150 F must be subtracted from the incore thermocouple readings to reflect the downcomer temperature.
This method will result in temperatures which will be lower than the 7
expected downtomer temperature. Thus, use of this methodology assumes o
that NDT will not be a problem.
. r BREAK SPECTRU'a AVE llAGE SYSTE: V010 FRACTION WITH Tile RC PU: PS OPERATIVE AND 2 HPi Pui.iPS 100
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V.
Guidelines for Non-LOCA Events Because of the broad spectrum of system conditions covered by the small break guidelines, the operator actions ard precautions identified to bring the plant to a long term cooling mode apply, in general, to any abnormal event which results in a decrease in RCS pressure. The small break guidelines will thus be utilized to update the emergency procedures for non-LOCA events; at a minimum, the following pertinent sections of'the small break guidelines.will be incorporated:
1.
RC Pump Trip Criteria and SG Level Control actions to promote natural circulation.
2.
RC pump Restart Criteria 3.
HPI Control Criteria 4.
The need to monitor system subcooling limits.
The items will be supplemented by the additional instructions / precautions to the effect that:
- l. For non-LOCA events, a restart of the RC pumps (1 per loop) and termination
.of SG fill is prudent to minimize system overcuoling due to addition of cold AFW to the OTSG's.
Note: The establishment of a subcooled condition (>50F) is a clean 1.1dication that a non-LOCA event or a LOCA for which a RCP trip is not required is not in progress.
- 2. HPI should be thr:ottled, when 50F subcooling is established, to avoid a
" pressurizer overfill.
- 3. During severe overcooling events, sufficient HPI water may be added, prior to achieving a.ubcooled condition (> SOF) and a pressurizer level (on-scale), such that the system may evolve to water solid state when the RC temperature recovers to a hot shutdown condition (s 530F).
e
1%
e 0perator action to control primary temperature (via secondary steam pressure control using the turbine bypass valves and/or atmospheric dum.ps) may be required to maintain pressurizer level on scale.
NOTE: The Operating Guidelines For Small Breaks have been modified to include Item 3 above.
l With operator training in the post-LOCA recovery methods in conjunction with modification of existing emergency procedures based on the small break guidelines, plant recovery and control can be achieved for any abnormal event for which an RCP trip is required.
J a
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REACTOR COOLANT PUMP TRIP ON HIGH PRESSURE INJECTION SIGNAL In response to NRC Bulletins79-05C and 79-06C, dated July 26, 1979, Florida Power Corporation wishes to respond to Short-term Action Item 1 A'on page.2 of 3.
This item requires that "upon reactor trip and initiation of HPI caused by low reactor coolant system pressure, immediately trip all operating reactor coolant pumps (RCP's)".
Florida Power Corporation proposes that a high pressure injection signal (HPI) be provided to immediately trip all operating RCP's.
The proposed design is outlined as follc.ws:
1.
A new ' Clark type PM relay and a solid state switch, Hamlin type P/N7, in parallel with existing relays 63Yl/RCI and 63Y2/RC1.
These relays and solid state switches are the same types that are presently used in the Engineered Safety Features Actuation System.
2.
The new components described above would be installed in both HPI ES Actuation "A" and "B" and operate by becoming de-energized when the 1500 psi bistable becomes de-energized below 1500 psi and is not bypassed.
3.
One normally closed contact from each of the relays in the HPI ES Actuation "A" and "B" would be seriesed and placed in the trip circuits of each of the RCP's.
The purpose of using contacts in series is to permit testing of a channel without tripping the
'RCP's.
~
4.
In addition, two (2) alarms would be added to the main control board to alert the operacor that (1) an HPI signal has tripped the RCP's or (2) sn alarm would alert the operator that one of the ES Actuation "A" or B" signals has been actuated.
5.
The schedule for implementation of this trip function is i-September 30, 1979.
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.-s E AC4A'.' ENTO f.'UNICi. At. t>TILITY crSIRlCT [] (201 S Street, B:x Mi33 Swa;.wr.to, Califstnia E.H3. (M6) U May 14, 1979 Mr. Harold R. Denton, Director Of fice of Nuclear Reactor Regulation U. S. Nuclear Regulatory Cornission Uashington, D. C.
20555 Rancho Seco Nucicar Cencrating Station Unit No. 1 Docket No. 50-312
Dear Mr. Denton:
The Sacramento Municipal Utility District is hereby submitting the current status of the actions proposed in our Ictter to you dated April 27, 1979, and required in the Co c.ission Order 7590-01 dated May 7, 1979.
.(a) 1.
Applicabic existing operating procedures were reviewed to assure timely and proper starting of notor driven Auxiliary Feedwater (AFW) pump (s) for vital AC buses upon loss of of fsite power.
The procedures reviewed includeo:
$)
A.51, Auxiliary, F< edwater System -
b)
D.1, Lond Rejection
. '. c)
- .L
~ -
n ;-
'.ha*d.'A.47, Condensate and'Feedwater Sy~ stem
'~
- 7.7
.-.E.
- m.:E.
' i. -[d).1 A.55, Main Feedwater_ Pumps System..
. b5-Og,-- J.5[' ' d- ".~
$ h,[..-
~
..y Load Rej.ect,f on,, was revised' a's. fo1Io...;
.. ;--.:.u. - -
ws:#
~r
~
'..,0f jhese procedures, D.1 4
~
a Section 5.4 now reads:
~
-.-. P -
.?
- 9 3., y.' ^
" Manually trip the Reactbr. if' the generator is not supplying i
power for the RCP's.
Verify Auxiliary Feedwater flow."
Section 6.5, NOTE:
"If nain feed pump op
.__ to A.51 section 7.5 t DUPLICATE DOCUMENT Euses 4A/4B.
Entire document previously entered into system under:
All licensed operators are bei b
assure their cognizance therco ANO
(
- < (
is being acconplished by means O
Training Supervisor (Attachmen No. of pages:
conpleted during the veck of M.
AN ELECTRIC SYST E M S E R YlN G
!.1 D R E THAN 6'0 0,0 0 0 IN THE H E A R 1 '0 F C A t i f 0 f *.