ML19208C637
| ML19208C637 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 08/10/1979 |
| From: | Hendrie J NRC COMMISSION (OCM) |
| To: | Matsui R HOUSE OF REP. |
| Shared Package | |
| ML19208C638 | List: |
| References | |
| NUDOCS 7909270167 | |
| Download: ML19208C637 (7) | |
Text
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e August 10, 1979 CHAIRMAN The Honorable Robert T. Matsui i
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Dear Congressman Matsui:
I apologize for not responding sooner to your letter of April 1,1979 but, as you know, the Commission and staff have been occupied with continued support of the efforts at Three Mile Island, orders to other facilities, appearances before the President's Commission and various committees of Congress that are investigating the Three Mile Island accident, and the realignment of our own resources and priorities to give immediate attention to the problems resulting from the accident.
Af ter the Three Mile Island accident, all holders of operating licenses for power reactors were instructed to take a number of immediate precautionary actions.
In addition, a series of acticos, both immediate and long term, were identified for all operating facilities using reactors designed by Babcock and Wilcox.
One of these is Rancho Seco, operated by the Sacramento Municipal Utilities District (SMUD).
In a letter to the NRC dated April 27, 1979, SMUD agreed to perform promptly the required immediate actions and to shut down Rancho Seco until these actions were completed.
The Commission issued a confirmatory order on May 7,1979, directing that Rancho Seco be maintained in a shutdown condition pending completion of the immediate (short-term) actions.
On June 21, 1979, the Commission ordered its Atomic Safety and Licensing Board Panel Chairman to set up a Board for the purpose of determining whether petitioners requesting a hearing on the May 7 Order met the Commission's personal interest test for participating in its proceedings.
If that test is determined to be met, the Board will then conduct any necessary hearing.
The June 21 Order stated that although the pendency of that proceeding would not preclude restart of Rancho Seco, the NRC Staff would, before permitting restart, brief the Commission as to its reasons for believing the short-term corrective actions to have been completed satisfactorily.
The Staff presented that briefing on June 26, 1979.
In a letter dated June 27, 1979, the Director of the Office of Nuclear Reactor Regulation notified SMUD of his finding that satisfactory compliance with the short-term requirements had been achieved, and that operation of the plant could therefore resume.
Enclosed is a copy of that letter and the attached safety evaluation.
Included in the changes are design modifications to some of the mechanical and electrical systems in the facility, revisions to operating and emergency procedures, and additional training for all licensed reactor and senior reactor operators.
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1050 174
The Honorable Robert T. Matsui Given below are answers to the specific questions you raised.
"In the Rancho Seco plant what is the possibility of voids being formed in the primary coolant system large enough to compromise the core cooling capability?"
Item (d) of the enclosed safety evaluation discusses the ability to main-tain the core cooling capability under conditions such as occurred at Three Mile Island.
The analyses confirm that a combination of heat removal systems presently available at Rancho Seco is capable of providing adequate core cooling.
A more detailed report is to be issued in the near future entitled, " Staff Report on the Generic Evaluation of Small Break Loss-of-Coolant Accident Behavior for Babcock and Wilcox Operating Plants" (NUREG-0565).
"Can operator action be depended upon to ensure continued core cooling in the event that such voids are formed? Furthermore, what assurances can be made that operators do not override automatic actions of engineered safety features without sufficient cause for doing so?"
Our analyses show that some operator action, both immediate and follow-up, is required under certain circumstances for a small-break accident and operator action must be relied upon. Guidelines developed by Babcock and Wilcox to assist the operating facilities in the development of emergency crocedures have been revised as a result of NRC review and have been imple-mented at Rancho Seco.
Extensive additional training for all licensed Operators has been conducted at Rancho Seco as a result of the Three Mile
- siand accident, including formal classroom and simulator training on design and procedural changes.
This training and the testing associated r;ith it have been audited by the NRC staff and is discussed further in
- tems (d) and (e) of the enclosed safety evaluation.
"The incident at Three Mile Island conclusively demonstrated the possibility of a Class 9 accident -- the release of radioactive material into the environment.
What steps are being taken by the NRC to include in the licensing process a review of the consequences of a Class 9 accident?"
The NRC has used the term " Class 9 accident" in referring to postulated accidents involving sequences of successive failures more severe than those
- ostulated for the design basis of protective systems and engineered safety fea tures.
A Class 9 accident is regarded primarily as one in which there is substantial core melting and breach of containment.
The TMI accident did not produce the massive radiation impacts that could be associated with postulated Class 9 accidents.
However, TMI demonstrated tnat multiple failures, each of which was individually taken into account in licensing, could result in a sequence of events which was not taken into account.
1050 175
p The Honorable Robert T. Matsui d ab The issue of the treatment of Class 9 accidents in the licensing process i:,
being considered by the NRC in a number of areas.
First, the task face on lessons learned from the Three Mile Island accident is considering the question from the viewpoint of reactor design and operation.
Second, a task force on siting issues is considering Class 9 accidents as they relate to site selection.
Third, NRC is reevaluating its emergency planning regulations to determine if any changes should be made with respect to major nuclear accidents.
The report of an NRC/ EPA joint task force on emergency planning, on which public comments have recently been received and analyzed by the NPC st ", is currently being considered by the Commission in connection with a poss' policy statement on the subject.
The joint task force's recommendations involve a planning basis that would reflect a wide range of conceivable accidents, including Class 9.
In addition, a specifi: case (Offshore Power Systems) involving the issue of whether certain environment considerations in connection with Class 9 accidents should be taken into account in issuance of a manufacturing license is currently before the Comission for adjudication.
"How can we be assured that undesired pumping of ra.ficactive liquids and gases will not occur inadvertently in the transfer of potentially radioactive gases and liquids out of containment?
In light of Three Mile Island, what protective modifications must be made to existing operating modes and procedures for all systems designed to transfer potentially ra oactive gases and liquids out of containtnent? What is the NRC's time.. :.ae for developing these modifications?"
These questions have been examined as part of the investigation of lessons learned from the accident at Three Mile Island. The structure that centains a nuclear reactor has mechanisms for isolating it from the environment.
In the TMI accident, this isolation occurred automatically, but not ur.til after water was pumped out of the containment by the automatic initio. ion of a transfer pump.
This water entered the radioactive was.te treatment system in the auxiliary building where some of it overflowed to the floor.
Although the present results from ongoing investigations indicate that this water was not the only source of offsite doses or high radicactivity levels in the auxiliary building, containment isolation is clearly an area where improvements should be made. The lessons learned task force in the Office of Nuclear Reactor Regulation has recommended the backfit of Section 6.2.4 on " Containment Isolation System" of the Standard Review Plan (copy enclosed) to all operating plants within the next few months to solve this problem.
"What alteration of plant reporting procedures for NRC notification is necessary to assure very early notification of serious events?"
Since the Three Mile Island accident, the NRC staff has taken several inrediate steps to improve licensee reporting and communication.
Dedicated telephone lines have been established between the NRC Operations Center in Betnesda, Maryland, and 68 of the 70 power reactors licensed for operation anc 14 licensed fuel cycle facilities.
(The other two reactors, Humboldt 1050 176
The Honorable Robert T. Matsui Bay and Indian Point Unit 1, have been shut down for more than a year.) The lines make it possible for operations personnel in those facilities to communicate. imediately and directly, in case of an incident or accident which could threaten the public health and safety, with members of the NRC Headquarters technical staff who are now on duty 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day, seven days a week. These lines are also tied into the regional offices of the NRC.
Also, to emphasize the need for prompt and continuous reporting, each power reactor licensee has been requested to review their prompt reporting procedures to assure that the NRC is notified within one hour of the time the reactor enters an uncontrolled or unexpected condition of operation.
Further, once notification has been made, an open continuous communication channel shall be established and maintained with the NRC.
The NRC staff is working on the problem of determining specifically the type of information that must be reported imediately.
"The incident at Three Mile Island has raised serious questions as to the adequacy and availability of its emergency storage capability of nuclear wastes.
Three Mile Island required the importation of lead bricks and holding tanks for surplus contaminated water.
If a similar accident occurred at Rancho Seco, is there safety equiipment on site to meet all reasonable emergencies?
If not, can the emetrgency equipment be made available within the appropriate response time?"
The provision of lead bricks and holding tanks during the TMI cooldown and recovery operaticn was r.ecessary to minimize exposures to workers and releases to the environ ent.
It was not emergency equipme'nt in the sense that it was necessary to control the reactor. This equipment is not available at Rancho Seco, but Rancho Seco does have a basin into which water containing low-level radioactivity could be discharged, if necessary, and where it could be held for treatrent or transport offsite.
If additional equipment was needed, it could be made available at Rancho Seco on approximately the same schedule as it was at TMI. Rancho Sece like all licensed reactors, is judged to have adequate engineered safety %
and design and operating features to cope with accidents.
"Given the projected difficulties in evacuating the moderately popu-lated region of Three Mile Island, can a realistic plan of evacuation be developed for the dense area surrounding the Rancho Seco plant? Are state and local authorities prepared for a Three Mile Island type accident? What should be the role of the federal government should local resources be inadequate?"
The U.S. Environmental Protection Agency published a report entitled
" Evacuation Risks -- An Evaluation" (EPA-520/6-74-002) in June 1974. A summary conclusion of tnat study was that "...large or small population groups can i>e effectively evacuated from impact areas with minimal death 1050 177
The Honorable Robert T. Matsui and injury risks and, in most cases, they can take care of themselves prnvided adequate plans are developed and executed to minimize potential problems that may occur peculiar to the impact area."
The recent fiRC/ EPA Task Force on Emergency Planning report " Planning Bas Response Plans in Support of Light Water Nuclear Pow 0396/ EPA 520/1-78-016)
Planning Zone (EPZ) for the plume exposure pathway of about 10 mile radius be established around light water nuclear power plants, like Rancho Seco.
before the EPA Administrator.The Task Force recommendatio Ta:k Force for this zone. Evacuation is one possible procective measure tha within this zone, or in parts of it, would depend upon the projected dent offsite consequences at the time.
Other protective measures that are of a thyroid blocking drug such as Potassium Iodide to radiciodine in the thyroid, should radioiodine be part of an airborne radiological release.
The area within a 10-mile radius of the Rancho Seco facility is sparsely populated.
The State of California has a radiological emergency response plan in which NRC has concurred.
Exercises using this plan have proven its operabil i ty.
local authorities, chiefly in sending qualified technical person appropriate equipment to augment the State and local government resources under the provision of the Interagency Radiological Assistance Plan (IPAP) managed by the Department of Energy.
As a result of the Three Mile Island accident, in which IRAP came into play, the Federal response role is being reevaluated with a view toward making needed improvements.
These answers to your questions are furnished in advance of completion of within and outside of the NRC concerning the Three Mile will address some of these concerns (copy attached).In additio be a great deal more to say on these matters in the future.There will undoubtedly this response will also satisfy the conterns expressed in your letter of I hope that June 27.
If you have any further questions, please do not hesitate to contact NRC.
Sincerely,,[
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_Jo_s_eph M. Hendrie
Enclosures:
1.
Ltr to J. J. Mattimoe from H. R. Denton dtd 6/27/79 2.
Standard Review Plan, Sec. 6.2.4,
" Containment Isolation System" 3.
Advance Notice of Proposed Rulemaking, 7/17/79
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June 27,1979 Docket No. 50-312 Mr. J. J. Mattimoe Assistant General Manager and Chief Engineer Sacramento Municipal Utlity District-6201 S Street P. O. Box 15830 Sacramento, California 95813
Dear Mr. Mattimoe:
By Order of May 7,1979, the Commission confirmed your undertaking a series of actions, bot', 'mmediate and long term, to increase the capability and reliability of the Rancho Seco Nuclear Generating Station to respond to various transient events.
In addition, the Order confirmed that you would shut down Rancho Seco on April 28, 1979, and maintain the plant in a shut-down condition until the following actions had been satisfactorily completed:
(a)
Upgrade the timeliness and reliability of relivery from the Auxiliary Feedwater System by carryi,ng out actions as identified in Enclosure 1 of your letter of April 27, 1979.
(b)
Develop and implement operating procedures for initiating and controlling auxiliary feedwater independent of Integrated Control System control.
(c)
Implement a hard-wired control-grade reactor trip that would be actuated on loss of main feedwater and/or turbine trip.
(d)
Complete analyses for potential small breaks and develop and implement operating instructions to define operator action.
(e)
Provide for one Senior Licensed Operator assigned to the control room who has had Three Mile Island Unit No. 2 (TMI-2) training on the B&W simulator.
By submittal of May 14, 1979, as supplemented by seven letters dated May 22, 24,29,30(3) and June 6,1979, you have documented the actions taken in response to the May 7 Order.
We have reviewed this submittal, and are satisfied that, with respect to Rancho Seco, you have satisfactorily completed the actions J'
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1050 179
Mr. J. J. Mattimoe.
prescribed in items (a) through (e) of paragraph (1) of Section IV of the Order, the specified analyses are acceptable, and the specified implementing procedures are appropriate.
The bases for these conclusions are set forth in the enclosed Safety Evaluation.
As noted on page 13 of the Safety Evaluaticr., you will be required to conduct a test during power operation to demonstrate operator capability to assume manual control of the Auxiliary Feedwater System independent of the Inte-grated Control System.
Appropriate Technical Specifications for Limiting Conditions for Operation and for surveillance requirements should be developed as so'on as practicable and provided to the staff within seven days with regard to the design and procedural changes which have been completed in compliance with the provisions '
of the May 7,1979 Commission Order.
The revised Technical Specifications should cover:
(1) Addition of flow indication to the Auxiliary Feedwater System; (2) Addition of the Anticipatory Reactor Trips; and (3) Changes in set points for high press,ure reactor trip and PORV actuation.
Within 30 days of receipt of this letter, you should provicle us with your schedule for completion of the long tem modifications described in Section II of the May 7 Order, and you should submit for staff review the model used in the analysis for potential small breaks referenced in yocr letter of May 14, 1979.
My finding of satisfactory compliance with the requirements of itecs (a) through (e) of paragraph (1) of Section IV of the Order will permit resumption of operation in accordance with the terms of the Commission's Order; it in no 5:ay affects your duty to continue in effect all of the above provisions of the Order pending your submission and approval by the Commission of the Technical Specifi-cation changes necessary for each of the required modifications.
Si ncerely, (P.
44 Harolo R. Denton, Director Office of Nuclear Reactor Regulation
Enclosures:
1.
Safety Evaluation 1nCn 10n 2.
Notice iUJU iUU cc w/ enclosures:
See next page
c' EVALUATION OF LICENSEE'S COMPLIANCE WITH THE NRC ORDER DATED MAY 7, 1979 SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION DOCKET NO. 50-312 June 27, 1979 1050 181
INTRODUCTION 1i p
%w By Order dated May 7,1979, (the Order) the Sacramento Municipal Utility District (SMUD or licensee) was directed by the NRC to take certain actions with respect to Rancho Seco Nuclear Generating Station.
Prior to this Order and as a result of a preliminary review of the Three Mile Island Unit No. 2 (THI-2) accident, the NRC staff initially identified several human errors that contributed significantly to the severity of the t: vent. All holders of operating licenses were s6sequently instructed to take a number of imediate actions to avo:o, mtition of these errors, in accordance with bulletins issued by the Commission's Office of Inspection and Enforcement (IE).
Subsequently, an additional bulletin was issued by IE which instructed holders of operating licenses for B&W designed reactors to take further actions, including immediate changes to decrease the reactor high pressure trip point and increase the pressurizer power-operated relief valve (PORV) setting.
- The NR staff identified certain other safety concerns that warranted additional short-term design and procedural changes at operating facilities having B&W designed reactors.
Those were identified as items (a) through (e) on page 1-7 of the " Office of Nuclear Reactor Regulation Status Report to the Commission" dated April 25, 1979.
After a series of discussions between the NRC staff and the licensee concerning possible design modifications and changes in cperr.ing procedures, the licensee agreed, in a letter dated April 27, 1979, to perform prcmptly certain actions.
The Commission found that operation of the plant
- [IE Bulletins Nos. 79-05 (April 1, 1979),79-05A (April 5, 1979), and 79-05B (Apri' 21, 1979) apply to all B&W facilities.]
1050 182
Phiu,udned-m should not be resumed until actions described in paragraphs (a) through (e) of paragraph (1) of Section IV of the Order were satisfactorily completed.
Our evalua' tion of the licensee's compliance with items (a) through (e) of paragraph (1) of Section IV of the Order is given below.
In performing this evaluation we have utilized additior.al laformation provided by the licensee on May 14, 22, 24, 29, 30, and June 6,1979, and numerous discussions with the licensee's staff.
Confirmation of design and procedure changes was made by members of the NRC staff at the Rancho Seco site.
An audit of the Rancho Seco reactor operators was also performed by the NRC staff to assure that the design and procedure changes were understood and were being correctly implemented by the operators.
EVALUATION Item a It was ordered that the licensee take the following action:
" Upgrade the timeliness and reliability of delivery from the Auxiliary Feedwater System by carrying out actions as identified in Enclosure 1 of the licensee's letter of April 27, 1979."
The Rancho Seco auxiliary feedwater (AFW) design has one turbine / motor tandem drive pump (P-318) that is automatically actuated and controlled independent 10sq 183
.g.
of offsite power, and one motor-driven AFW pump (P-319) that is automatically started,.but must be manually transfarred to a vital AC bus if offsite power is lost.
The turbine / motor driven pump will be manually started, according to procedure, from a vital AC bus if the turbine drive fails.
By reference above to Enclosure (1) of the licensee's letter of April 27, 1979, it was ordered that the licensee:
"1.
Review procedures, revise as necessary and conduct training to ensure timely and proper starting of motor driven aaxiliary feed-water (AFW) pump (s) from vital AC buses upon loss of offsite power."
The licensee has developed Section 7.5 of Operati ng Procedure A.51 (" Auxiliary Feedwater System") to provide specific direction for the operator on the steps required to load motor driven pump P-319 on nuclear service bus 4A and to secure the steam to the turbine on the dual-drive pumo P-318, in the event of inoperability of the steam drive, and load the motor drive on nuclear service bus 4B.
Bypass keys are required to complete the connection of the auxiliary feedwater pump motors to the diesel powered buses (nuclear service buses 4A and 4B); these keys are available in the office adjacent to the control room.
Emergency Procedure 0.1 (" Load Rejection") directs the operator to use Operating Procedure A.51 if main feed pump operation cannot be maintained.
The NRC staff verified that the operators are knowledgeable in the procedure for loading the AFW pumps on the vital AC buses.
The NRC staff concludes that the licensee has adequate procedures and the operators are trained to start the AFW system from diesel powered buses upon loss of offsite power or load rejection and therefore, is in compliance with this part of the Order.
1050 184 3
It was also ordered that:
"2.
To assure that AFW will be aligned in a timely manner to inject on all AFW demand events when in the surveillance test mode, procedures will be implemented and training conducted to provide an operator at the necessary valves in phone communications with the control room during the surveillance mode to carry out the valve alignment changes upon AFW demand events."
Surveillance Procedures SP 210.01A and SP 210.01B are used for the quarterly surveillance and inservice testing of auxiliary feed pumps P-318 and P-319, respectively.
These procedures have been revised to include the following statement; " Station an operator at FWS-055, auxiliary feedwater system full flow recirculation vrive in continuous communication with the control room until FWS-055 is secured closed at the completion of this test."
In addition to the above procedure revisions, the licensee has added FWS-492 (bypass valve for FWS-055) to the " Locked Valve List" (SP 214.03).
The licensee has also incorporated independent verification of valve lineups following surveillance testing and/or maintenance of the AFW system.
The NRC staff has reviewed SP 210.01A and SP 210.01B to verify that the procedures contain specific directions to return each valve that was operated during the conduct of the surveillance test to its proper position.
The local operator has to close a valve (FdS-855) when so instructed by the control room operator 1050 185
_4
or if he loses communication with the control room.
The Nnc staff has verified that the operators are familiar with this test procedure.
We conclude that the licensee has adequate procedures to assure that AFW will be aligned in a timely manner to inject on all AFW demand events when in the surveillance test mode and therefore, is in compliance with this part of the Order.
It was ordered that:
"3.
Procedures will be developed and implemerited and training conducted to provide for control of steam generator level by use of safety grad.e AFW bypass valves in the event that ICS stearn generator level control fails."
The licensee has developed Emergency Procedure D.14 (" Loss o-f Steam Generator Feed") that describes the symptoms that would result from a lloss of main feedwater control that may have been caused by an integrated control system (ICS) failure.
The procedure has been reviewed by the NRC staff.
The operator is directed to restore feedwater to the steam generators by one cf three methods.
The preferred method is described in Section 7.7 of Operating Procedure A.51 (" Auxiliary Feedwater System").
Section 7.7 directs the operator to:
close the ICS controlled AFW control valves; start the AFW pumps; and maintain the steam generator levels, specified in the procedure, by manually operating the motor driven AFW bypass valves from the control room.
In this mode the pumps and valves will operate independent of the ICS.
The operator is provided with AFW flow rate and steam generator level indications in the control room for each steam generator.
1050 186 3
P00R ORIGINAL Since the AFW bypass valves will fully open on a safety features actuation signal (SFAS)*, the operator is provided with instructions on how to take manual control of the valves after a SFAS.
NRC staff has conducted an audit of the operator training and verified that the operators have been trained to cerry out' those procedures.
The NRC staff concludes that the licensee has developed adequate procedures and operator training to control AFW flow to the steam generators to specified values independent of the ICS, should a failure of the IES occur, and therefore,,
is in compliance with this part of the Order.
It was also ordered:
"4.
Verification that Technical Specificat' ion reodrements of AFW capacity are in accordance with the accident analysis will be conducted.
Pump capacity with mini flow in service wil' M so be verified."
The licensee has conducted the verification that Technical Specifica'. ion requirements of AFW capacity are in accordance with the accident analysis for the Rancho Seco Nuclear Station.
The Technical Specificaticin states, as a
- [The safety features actuation system (SFAS) monitors variables to detect loss of reactor coolant system boundary iatagrity.
Upon det.ection of "out-of-limit" conditions of these variables, it initiates ernergency core cooling (ECC) which consists of high pressure injection (HPI) and low pressure injection (LPI), Reactor Building cooling and isolation, and Reactor Building spray systems.
Additionally, it starts diesel generators GEA and GEB, which are in standby redundance with the nuclear service buses 4A and 4B.]
1050 187
.g
limiting condition for operation, capability to supply feedsater at a process flow rate corresponding to a decay heat level of 4.5 percent of full reactor power from at least one of the following means:
(a) a condensate pump and a main feed pump, or (b) a condensate pump, or (c) an auxiliary feedwater pump.
A letter from Babcock & Wilcox to the licensee, dated May 16, 1979, states that it has performed an analysis of the required AFW flow rate for the Rancho Seco Plant which shows that a decay heat level of 4.5 percent of rated power, plus the heat input from the RCPs, will require a total flow rate to either or both steam generators of approximately 760 gpm.
r Each of the two AFW pumps are sized to deliver 780 gpm to steam generators with 60 gpm mini flow in service.
This pump capacity exceeds the minimum required AFW flow rate in the Rancho Seco safety analysis and Technical Specifi-cations.
AFW pump capacity, with mini flow in service, has been verified by performing the quarterly "AFW System Surveillance Test" and the " Auxiliary Feedsater riow Indicator Functional Test" (STP 612).
The results of these tests demonstrated that each of the two AFW pumps has the capability to deliver a minimum of 780 gpm into the steam generators, with mini flow in service.
The licensee will reconfirm the minimum AFW flow rate to the steam generators in a test immediately following startup.
1050 188 7
Based on our review of the AFW flow rate test results, performed to date, we c.,nclude that the licensee is in compliance with this part of the Order.
It was also ordered that:
"5.
Modifications will be made to provide verification in the control room of AFW flow to each steam generator."
To verify that AFW is being pumped to the steam generators, the licensee has installed Clampitron Flowmeters on both of the AFW injection flow paths, downstream of the AFW control valves, so that the actual flow rate te each steam generator will be measured.
The Clampitron Flowmeters consists of transducers, attached to the AFW piping, connected to a flow display computer.
On command from the flow display computer, the trdnsducer transmits an ultra-sonic beam through the water inside the pipe and the velocity of the beam, as affected by AFW flow, is analyzed by the flow display computer, which calculates the AFW flow rate in gpm.
The AFW flow rate is displayed in the control room.
A calibration test (STP-612) was conducted by the licensee to functionally test the performance of the flowmeters.
Performance of this test demons trated that the indicated flow rate agreed with the calculated flow rate within the 20% acceptance criteria specified in the procedure.
Based on our review of this design modification and test results, we conclude that the licensee is in compliance with this part of the Order.
1050 189
.g.
It was also ordered that the licensee:
"6.
Review and revise, as necessary, the procedures and training for providing alternate sources of water to the suction of the AFW pumps."
Control room alarms are available to alert the operator to perform the manual transfer of the AFW supply source from the condensate storage tank (CST) to the plant reservoir.
The CST is designed to seismic Category I criteria.
The licensee has reviewed and revised his Emergency Procedures D.10 (" Loss of Reacter Coolant Flow /RCP Trip"), 0.14 (" Loss of Steam Generator Feed"), and Operating Procedure A.51 (" Auxiliary Feedsater System") to provide guidance for the operator to obtain an alternate source of water for the s_ction of the AFa' pumps.
Tne revised procedures' require the op'erator to break condenser vaccur when the level reaches a level alarm point of approxirately 29 feet and tc shift the AFW pump suction to the plant reservoir when the CST level is dcan to a second alarm point of approximately 3 feet frc the bottom of the tank.
The capacity of the CST is large enough to provide cocling for about 24 nours before this transfer is required.
The shifting to an alternate source of AFV pump suction is accomplished by manually operating four isolation valves at a local valve station.
The operator has about 40 minutes to effect the transfer.
The NRC staff has reviewed the revised Emergency Procedures D.10 anc D.14 and Operating Procedure A.51 and concludes that thesa procedures provide sufficient guidance to the operator for a timely shifting to an alternate water source for the AFW pumps, before the CST is emptied.
1050 190
_g.
The NRC staff has verified that the control room operators are properly trained to carry out these procedures.
We conclude that the licensee has complied with the requirements of this part of the Order.
It was also ordered that:
"7.
Design review and modification, as necessary, will be conducted to provide control room annunciation for all auto start conditions of the AFW system."
The licensee has provided indication for all auto start ccnditions of the AFW system on an annunciation panel inside the control room.
The conditions which will actuate the annunciator are:
1 (a) loss of all reactor coolant pumps, or (b) low discharge pressure (850 psig) on both main feedsater pumps, or (c) manual start of the motor driven AFW pump.
A safety features actuation signal, which will also automatically start AFW, had already been annunciated in the control room before the current modifications.
Based on our review of this design modification, we conclude that the licensee is in compliance with this part of the Order.
It was ordered that:
1050 19I L
ti-
"8.
Procedures will be deveioped and implemented and training conducted
~
.to provide guidance for timely operator verification of any automatic initiation of AFW."
The conditions that will' automatically initiate auxiliary feedwater are adequately described in Operating Procedure A.51, (" Auxiliary Feedwater System").
The operators are directed, as an immediate action, to verify that the AFW flow has automatically started on loss of both main feedwater pumps in Emergency Procedure D.14 (" Loss of Steam Generator Feed") and on loss of all reactor coolant pumps in Emergency Procedure D.10 (" Loss of Reactor Coolant Flow /RCP Trip").
Both procedures require the following immediate actions by the operator:
verify that the auxiliary feedwater pumps have cutomatically started; that there is flow to the steam generators; and that the proper steam generator levels are being maintained.
The NRC staff has performed an audit and verified that the operators are trained in these procedures.
Based on review of these procedures, we conclude that the licensee has providec.
guidance for timely operator verification of any automatic initiation of AFW and therefore, is in compliance with this part of the Order.
It was also ordered:
"9.
Verification will be made that the air operated level control valves (a) Fail to the 50% open position upon loss of electrical power to 1050 192.
the electrical to pressure converter, and (b) Fail to the 100% open position upon loss of service air.
The AFW bypass valves are safety grade."
,k The licensee has completed its verification test for the failure mode of the air operated level control valves.
The test results show that both air operated level control valves fail to the 100% open position on loss of air pressure at the valve operators.
On tests for loss of ':ontrol signal to the electric to pressure converters, one level control valve failed to the 50% open position and the other one failed to the 60% open position, which are acceptable.
The AFW bypass valves are safety grade, motor-operated valves which are operated independently from the ICS as discussed in Part 3 above.
Based on our review of the test results on the air operated level control valves and the safety grade design of the bypass valves, we conclude that the licensee is in compliance with this part of the Order.
Based upon our evaluation, we conclude that the licensee has upgraded the timeliness and reliability of delivery from the AFW system by carrying out the actions identified in Enclosure 1 of the licensee's letter of April 27, '1979, and therefore, is in compliance with Item (a) of the Order.
Item (b)
It was ordered that the licensee:
1050 193 _ _ _
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" Develop and implement operating procecures for initiating and controlling auxiliary feedwater independent of Integrated Control System (ICS) control."
We have reviewed the revised procedures for the AFW system to assure that there is sbfficient guidance for the operator to actuate the system if the automatic initiation failed, and to control steam generator levels at the required values.
The review of the procedures focused on verifying that the operator is directed to observe the proper instruments and that the operator is directed to maintain specific values of parameters by manual control, such as steam generator levels.
The review also determined that the cperator should confirm the validity of the instrument readings of certain key parameters, such as steam generator levels.
The necessary modifications to the procedures to satisfy these requirements were presented to the licensee, and the NRC staf f has verified that the modifications nave be'en incorporated in the procedures.
(See further discussion of these procedures in part 3 of Item (a.)
The licensee will conduct a startup test at low power (<1%) to demonstrate the capabihty to provide and control flow to the steam generators using the AFW bypass valves.
During the visit to the site, the NRC staff walked through the AFW procedures with the operators to evaluate whether the procedures were functionally adequate.
In addition, the NRC staff audited a sample of Rancho Seco operators to determine if they were familiar with the revised procedures and couT d implement them correctly.
Based on the NRC staff audit, we conclude that. the revised procedures and cperator training are satisfactory and therefore, the licensee is in compliance with Item (b) of the Orcer.
1050 194 Item (c)
The original Rancho Seco design did not have any direct reactor trips that would be initiated by a malfunction in the secondary system.
To obtain an anticipatory reactor trip (rather than delaying the trip until a prima system parameter exceeded its trip setting) the licensee committed to install a hard-wired, control grade, reactor trip on loss of all main feedwater and/or turbine trip.
The Order requires that the 1in nsee:
" Implement a hard-wired control grade reactor trip that would be actuated on loss of main feedwater and/or turbine trip."
The licensee has added contro' grade circuitry to Rancho Seco, which is designed to provide an automatic reactor trip when either the main turbine trips or all main feedwater is lost.
The purpose of the anticipatory trip is to minimize the potential for opening of the power-operated relief valve (PORV) and/or the safety valves or the pressurizer.
The licensee has indicated that this new circuitry meets this objective by providing a reactor trip during the incipient stage of the related transients (turbine trip and/or loss of main feedwater).
The main turbine trip is sensed by an existing, normally deenergized relay in the main turbine / generator protection system.
The relay is energized by the protective trips of the turbine and/or generator.
Power is supplied by an onsite battery source.
Tne loss of all main feedwater is sensed by two newly installed pressure switches (one in each of the two main feedwater pump discharge lines).
The 1050 195
pressure switches actuate (close) on low pressure in the header.
Power is supplied by the same or. site battery source.
In order to prevent an inadvertent reactor trip during startup or shutdown, the loss of all main feedwater trip input is cut-out of the circuitry by a keylock switch.
The key for this switch is maintained in the custody of the shift supervisor and is located in the centrol room.
When the switch is placed in the " cut-out" position, it is annunciated on the main control board.
The operating procedures specify when the switch is placed in the " normal" or " cut-out" position.
Either signal (turbine trip or loss of all main feedwater) will actuate a reactor trip relay, which in turn provides an input to both of the st.unt coils of the AC reactor trip breakers.
Energizing both of the shunt coils caus:cs a reactor trip.
r The licensee has analyzed this additional circuitry with respect to its independence from the existing reactor trip system.
They have stated that the shunt coil is part of the existing AC reactor trip breaker.
Each shunt coil is powered by a separate Class IE 125 VDC supply and operates independently from the 120 VAC undervoltage trip coil which receives the safety grade reactor trip signal.
An NRC inspector has confirmed that the check-out tests for this circuitry have been completed successfully.
In addition, the licensee has committed to perform a monthly periodic test on the added circuitry in order to demonstrate its ability to open the AC reactor trip breakers via the shunt coil.
-lE-1050 196
Based on our review of the implementation of the trip circuitry, with respect to its ir. dependence from the existing reactor trip circuitry, we conclude that this addition will not degrade the existing reactor protection system design.
Based on the licensee's design modifications and commitment to perform a mcnthly test on the new circuitry, we conclude that there is reasonable assurance that the systera will perform its function.
On the basis of tne evaluation at.ove, we conclude that the licensee has complied with the requirements of Item (c) of the Order.
Item (d)
This item in the Order requires the licensee to:-
" Complete analyses for potential small breaks and develop and implement operatir.g instructions to define operator action."
In the licensee's letter of April 27, 1979, the licensee committed to providing the analyses and operating procedures of this requirement.
Babcock and Wilcox, the reactor vendor for the Rancho Seco plant, submitted analys es entitled, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant" and supplements to these analyses (References 1 through 6).
The major parameters used in this generic study 1050 197
bound the Rancho Seco plant.
The staff evaluation of the B&W generic study has been completed and the results of the evaluation w;ni be issued as a NUREG report in June 1979.
h b.!MM A principal finding of our generic review is a reconfirmation that Loss-of-Coolant Accident (LOCA) analyses of breaks at the lower end of the small break spectrum (smaller than 0.04 sq. ft.) demonstrate that a combination of heat removal by the steam generators, the high pressure injection system and operator action ensure adequate core cooling.
The AFW system used to remove heat through the steam generators has been modified to enhance its reliability as discussed in item (a).
The high pressure injection system is capable of providing emergency core cooling even at the safety valve pressure setpoint.
The ability to remove heat via the steam generators has always been recognized to be an important consideration when analyzing very smaT1 breaks.
Separate sensitivity analyses were performed assuming permanent loss of all feedwater (with operator initiation of the high pressure injection system at 20 minutes) and loss of feedwater for only the first 20 minutes of the accident for breaks of 0.01 sq.
ft.
Reactor core uncovery is not predicted for these events.
The calculated peak cladding temperature was less than 800 F, well below the 10 CFR 50.46 requirement of 2200 F.
These results are applicable to Rancho Seco considering the ability to manually start the redundant AFW pumps from the control room, assuming failure of automatic AFW actuation.
Another aspect of the study was the assessment of recent design changes on the lift frequency of pressurizer safety and relief valves.
The design changes 1050 198 included:
a change in the setpoint of the pressurizer power-operated relief valve (PORV) from 2255 psi 1.o 2450 psi; change in the high pressure reactor trip setpoint from 2355 psi to 2300 psi; and the installation of an anticipa-tory reactor trip on turbine trip and/or on loss of all main feedwater.
In the past,'during the turbine trip or loss of feedwater transients, the PORV lifted.
With the design changes the initial pressure increase of these tran-sients do not result in lifting of this valve.
However, the consequent depressurization could initiate safety injection which in turn could repres-surize the system and lift the relief valve.
It is expected that thr. operator would terminate HPI before the relief valve or safety valves lift, since the 50 F subcooling criteria would be satisfied at pressures below the PORV setpoint.
Also, lif ting of both the PORV and safety valves might occur in the case of control red withdrawal or inadvertent boron dilution -transients, using the normally conservative assumptions found in the Chapter 15' safety analyses.
The above design changes do not effect the lift frequency of 'the valves for these Chapter 15 safety analyses.
Based en our review of the small break analyses presented by B&W, the staff has determined that a loss of all main feedwater with (a) an isolated PORV, but safety valves opening and closing as designed, or (b) a stuck open PORV does not result in core uncovery, provided either AFW or 2 HPI pumps is initiated within 20 minutts.
Based on the acceptable consequences calculated for small break LOCAs and loss of all main feedwater events coupled with the expected reliability of the AFW and HPI systems, we conclude that the licensee has complied with the analyses portion of Item (d) of the Order.
}03)
To support longer term operation of the facility, requirements will be developed for additional and more detailed analyses of loss of feedwater and other anticipated transients.
More detailed analyses of small break LOCA events are also needed for this purpose.
Accordingly, the licensee will be required to provide the analyses discussed in Section 8.4.1 and 8.4.2 of the recent NRC
" Staff Report of the Generic Assessment of Feedwater Transients in Pressurized Water Reactors Designed by the Babcock and Wilcox Company" (NUREG 0560).
Further details on these analyses and their applicability to other PWRs and BWRs will be specified by the staff in the near future.
In additiori, to assist the staff in developing more detailed guidance on design requirements of relief and safety valve reliability during anticipated transients, as discussed in Section 8.4.6 of the NUREG report, the licensee will be required to provide analyses of the lift frequency and mechanical reliability of the pressurizer relief and safety valves of the Rancdo Seco facility.
The B&W analyses show that some operator action, both immediate and -followup, is required under certain circumstances for a small break accident.
Immediate operator action is defined as those actions committed to memory by Ihe operators which must be carried out as soon as the problem is diagnosed.
Foll ow-up actions require operators to consult and follow the steps in writteri and approved procedures.
These procedures must always be readily available in the control ioom for the operators' use.
Guidelines were developed by B&W to assist the operating B&W facilities in the development of emergency procedures for the small break accident.
1050 200 3
P00R ORIGINA The " Operating Guidelines for Small Breaks" were issued by B&W on May 5, 1979 and reviewed by the !!RC staff.
Revisions recommended by the staff were incorporated in the guidelines.
In response to these guidelines, the staff at Rancho Seco made substantial revisions to Emergency Procedure D.5 (" Loss of Reactor Coolant / Reactor Coolant System Pressure") and Operating Procedure B.4
(" Plant Shutdown and Cooldown").
These procedures define the required operator action in response to a spectrum of break sizes for a loss-of-coolant accident in conjunction with various equipment availability and failures.
Emergency Procedure 0.5 (EP D.5) is divided into three sections.
The first section deals with a small leak within the capability of a makeup pulp.
In this case, the operators proceed with an orderly plant sh/JtdoWn unless pres-surizer or makeup tank levels f all below prescribed limit.s.
If these limits are exceeded tr.9 reactor is manually tripped andshigh pressure injection is initiated.
The second section of EP D.5 defines the required operate >r action fo: a small break not witnin the capability of a makeup pump.
This section provides the operator with the guidance necessary to achieve a safe hot shutdown condition for a variety of degraded conditions.
If all feedwater is lost, a heat removal path is established by the high pressure injecticn system through the break and the pre 3surizer power-operated relief valve or the safety valves.
Once feedwater is reestablished, the steam generators can be used as a heat sink.
If the reactor coolant pumps are not available, the operator is directed to Operating Procedure B.4 (OP B.4) which defines the actions necessary to cool down the plant by natural circulation.
Additional guidance is provided in OP E.4 if natural circulation is not immediately achieved.
1050 201
_ u,.
The third section of EP D.5 defines the actions necessary in the event of a large rupture.
In this case the system depressurizes to the point of low pressure injection.
For all cases in which high pressure injection is manually or automatically initiated, the operators are specifically instructed in EP D.5 to maintain maximum HPI flow unless one of the following criteria are met:
(1) The LPI system is in operation and providing cooling at a rate in excess of 1000 gpm and the situation has been st.able for 20 minut2s, or (2) All hot and cold leg temperatures are at least 50 degrees below the saturation temperature for the existing RCS pressure.
If the 50 degrees subcooling cannot be maintained after HPI cutoff, H?I shall be reactuated.
A requirement to determine and maintain 50 F subcooling has been incorporated in all other procedures in which HPI has been manually or automatically initiated.
These procedures include, " Steam Supply System Rupture," and " Loss of Steam Generator Feedwater."
Each of these procedurn, in addit. ion to the " Loss of Reactor Coolant / Reactor Coolant System Pressure" procedure, provide additional instructions to the operators in the event of faulty or Inisleading indications. 1050 202
P0 Q D E NAL A subsequent action statement directs the operators to check alternate instru-mentation channels to confirm key parameter readings.
The Rancho Seco staff has made revisions to all of their emergency procedures to include this requirement.
If feedwater is not initially available following a transient or accident, core cooling is maintained by flow from two HPI pumps and relief through the PORV, which is opened by the operator.
B&W has performed studies that show that density differences between the downcomer and reactor core will cause recirculation flow between the core exit and downcomer via the vent valves.
Mixing of the hot core exit water with the cold HPI water will provide suf fi-ciently warm vessel temperatures to preclude any significant thermal shock effects to the vessel.
Under these conditions with no circulation of water from the steam generators, the cold leg thermocouple (located upstream of the reactor coolant pump) does not provide a satisfactory indication of the vessel temperature.
B&W has recommended using the core exit thermocouples as a measure of vessel temperature, based on B&W analyses that conservatively show that the vent valves will open at temperature differences between the core exit and downcomer of less than 150 F.
They have also proposed a more appropriate pressure-temperature limit curve for the vessel that reflects allowable stresses under these faulted :onditions (no feedwater).
The NRC staff has reviewed these guidelines and finds them acceptable because of the expected recirculation through the vent valves and the vessel stress limits used.
The licensee has incorporated these revised guidelines in his procedures for loss of all feedwater.
Subsequent restoration of AFW would depressurize the reactor coolant system to below 600 psi where pressure vessel 1050 203 22-
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~
integrity is assured for any reasonable thermal transients that might subsequently occur. We conclude that further reliability analyses are needed as part of the long-term requirements of the Order to confirm that AFW can be restored (if lost) in a reasonable period of time.
B&W has agreed to provide a detaile'd thermal-mechanical report on the behavior of vessel materials for these extreme conditions, to be applicable generically to the Oconee class of plants, which includes Rancho Seco.
The " Loss of Reactor Coolant / Reactor Coolant System Pressure" procedure was reviewed by the NRC staff to determine its conformance wit:h the B&W guicelines.
Comments generated as a result of this review were incorporated in a further revision to the procedure.
A member of the NRC staff walited through this emergency procedure in the Rancho Seco control room.
The procedure was judged to provide adequate guidance to the operators to' cope with a small break loss-of-coolant accident.
The instrumentation necessary to diagnose the break, the indications and controls required by the action statements, and the administrative controls which prevent unacceptable limits from being exceeded are readily available to the operators.
We conclude that the operators should be able to use this procedure to bring the plant to a safe shutdown condition in the event of a small break accident.
An audit of seven of 14 licensed operators and senior operators assigned to shift duty (22 total licensed personnel) was conducted by the NRC staf f to determine the operators' understanding of the small break accident, including how they are required to diagnose and respond to it.
The Rancho Seco staff has conducted special training sessions for the operators on the concept and-1050 204 use of EP D.5.
The audit revealed that, except for one deficiency, the operators had sufficient knowledge of the small break phenomenon and the requirements of the procedure.
This deficiency, verification of natural circulation, was brought to the attention of the plant staff.
Each licensed individual received additional training in this area by the plant training organization and General Physics Corporation.
They also received training on the revisions made to EP D.5 as a result of the NRC review.
This additional training has been completed and verified by the NRC staff.
The audit of the operators also included questioning about the THI-2 incident and the resulting design changes made at Rancho Seco.
The discussions covered the initiating events of the incident, the rest - se of the plant to the simul-taneous loss of feedwater and small break LOCA (PORV stuck open), and the operational actions that were taken during the cdurse of the incident.
We identified a deficiency in interpreting the initial sequence of the TMI-2 incident on the part of several of the operators.
Additional training has been conducted in this area by the plant staff and their consultant and verified by the NRC staff.
Otherwise, we found their level of understanding suf ficient to be able to respond to a similar situation if it happened at Rancho Seco.
We also concluded they have adequate knowledge of subcooling and saturated conditions and are able to recognize each in the primary coolant system by various methods.
The AFW system was also discussed during the audit to determine the operators' ability to assure proper starting and operation of the system during normal conditions, as well as during adverse conditions such as loss cf 1053 205 P00R DEiD1 of fsite power or loss of normal feedwater.
The long term cperation of the system was examined to evaluate the operators' ability to use available. mual controls and water supplies.
The level of understanding was found to be sufficient to assure proper short and long term AFW flow to the steam generators.
In addition to the oral audit conducted by the NRC, the licensee administered a written examination to all licensed personnel.
Individuals scoring less than 90 percent on the exam will receive additional training and will not assume licensed duties until a score of at least 90 percent is attained on an equivalent, but different exam.
The written exam and the grading was audited by the NRC staff and judged to be satisfactory.
The staff will also review all subsequent results and records as part of the normal inspection function of the Rancho Seco requalification program.
We conclude that there is adequate assurance that the operators at Rancho Seco haverand will continue to receive a high level of training concerning the TMI-2 accident and the consequent impact at their station.
Based on the foregoing evaluation, we conclude that the licensee has complied with the requirements of Item (d) of the Order.
Item (e)
The Order requires that the licensee:
" Provide for one senior licensed operator assigned to the control room who has had TMI-2 training on the B&W simulator." I
The licensee has confirmed that this item of the Order has bia1 co=pleted and has further co==itted that all reactor operators and senior reactor operators vill have co=pleted the TMI-2 simulator training at B&W by June 21, 1979.*
This training consists of a class discussion of the TMI-2 event followed by a demonstration of the event on the si=ulator as it occurred and the proper actions that should be taken to control the accident.
The class discussion is about four hours long and the remainder of the session is conducted on the simulator. The M I-2 event, including operational errors, is demonstrated to each operator. The event is again initiated and the operators are given
" hands-on" experience in successfully regaining control of the plant by several methods.
Other transients which result in depressurization and saturation conditions are presented to the operators and they cust =aneuver the plant to a stable, subcooled condition.
9 Based on the above actions by the licensee, we conclude that the licensee is in co=pliance with Ite= (e) of the Order.
Conclusion We conclude that the actions described above fulfill the requirements of our Order of May 7,1979 in regard to Paragraph (1) of Section IV.
The licensee having met the requirements cf Paragraph (1) may restart Rancho Seco as provided by Paragraph (2).
Paragraph (3) of Section IV of the Order re=ains in force until the long terr modifications set forth in Section II of the Order are completed and approved by the NRC.
This action has been co=pleted and satisfies *he long-ter= porticrn of the Order in this regard. 1050 207
REFERENCES 1.
Letter from J. H. Taylor (B&W) to R. J. Mattson (NRC) transmitting report entitled, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant," dated May 7,1979.
2.
Letter from J. H. Taylor (B&W) to R. J. Mattson (NRC) transmitting revised Appendix 1, " Natural Circulation in B&W Operating Plants (Revision 1),"
dated May 8, 1979.
3.
Letter from J. H. Taylor (B&W) to R. J. Mattson (NRC) transmitting additional information regarding Appendix 2, " Steam Generator Tube Thermal Stress Evaluation," to report identified in Item ? above, dated May 10, 1979.
o 4.
Letter from J. H. Taylor (B&W) to R. J. Mattson (NRC), providing an analysis for "Small Break in the Pressurizer (PORV) with no f.uxiliary Feedwater and Single Failure of the ECCS," identified as Supplements 1 and 2 to Section 6.0 of report in Item 2, dated May 12, 1979.
5.
Letter f rom J. H. Taylor (B&W) to R. J. Mattson (NRC), pr oviding an analysis for "Small Break in the Pressur'Ter (PORV) with no Auxiliary Feedwater and Sii.;1e Failure of the ECCS" identified as Supplements 1 and 2 to Section 6.0 of repert in Item 2, dated May 12, 1979.
1050 208 _. _
s 6.
Letter from J. H. Taylor (B&W) to R. J. Mattson (NRC), providing Supplement 3 to Section 6 of report in Item 2, dated May 24, 1979, r
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NUREG 75/Os7 pa ese U.S. NUCLEAR REGULATORY COMMISSION f ~,
g STANDARD REVIEW PLAN i
1
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OFFICE OF NUCLEAR REACTOR REGULATION SECTION 6.2.4 CONTAINMENT ISOLATION SYSTEM
=
REVIEV RESPONSIBILITIES Primary - Containment Systems Branch (CSB)
[
Secondary - Accident Analysis Branch (AAB)
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Instrtssentation and Control System Branch (ICSB) g E
Mechanical Engineering Branch (MEB)
E Structural Engineering Branch (SEB)
-r Reactor Systems Branch (RSB)
Power Sysues Branch (PSB)
I.
AREAS OF REVIEW The design objective of the containment isolation system is to allow the norsal or emer-gency passage of fluids through the containment boundary while preservir.ig the ability of the boundary to prevent or limit the escape of fission products that say result from postulated accidents. This SRP section, therefore, is concerned with tne isolation of fluid systems which penetrate the containment boundary, including the oesign and testing requirements for isolation barriers and actuators. Isolation barriers include valves, closed piping systems, and blind flanges.
The CSB reviews the information presented in the applicant's safety anakysis report (SAR),
regarding containment isolation provisions to assure conformance with the requirements of Ganeral Design Criteria 54, 55, 56 and 57. The CSB review covers the fcillowing aspects
~
of containment isolation:
i.
The design of containment isolation provisions, including:
The nummer and location of isolation valves, i.e., the isolation valve arrange-a.
ments and the physical location of isolation valves with respect to the containment.
b.
The actuation and control features for isolation valves.
'c.
The positions of isolation valves for normal plant operating enditions (includ-
~.
ing shutdown), post-accident conditions, and in the event of valve operator
=
power failures.
~
d.
The valve actuation signals.
The basis for selection of closu DUPLICATE DOCUMENT
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Entire document previously
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USNRC STAND AR f* '.~~.~. O"* W'".
entered into system under:
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ANO No. of pages:
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._r 1050 210