ML18081B123
| ML18081B123 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 12/31/1979 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Librizzi F Public Service Enterprise Group |
| References | |
| IEB-79-06A, IEB-79-6A, NUDOCS 8002280017 | |
| Download: ML18081B123 (17) | |
Text
December 31,
- 1979 Docket No.: 50-272 Mr. F. P. Librizzi, General Mnnager Electric.Production Public Service El ~ctri c & Gas Con1pany BO fark Place~ Room 7221 Newark 11 New Jersey 07101*
Dear Mr. Librizzi:
TERA TIC D Eisenhut
- B Grimes RVollmer T J Varter.
W Russell OELD NSIC, J R NRR Rdg Buchanan I&E (.3)
SB Rdg & 'chron ORB#l ROG Project Manager Licensing Assi~tant A.
Schw~ncer Jordan, I&E
.. W F Kane D F Ross p 0 I R~i lly IAf,.l. Re~>
SUBJECT:
.* NRC STP.FF EVALUATION OF PSE&G RESPONSES TO IE **suLL~TINS 79-06i\\
AND 79-06A~ REVISION l, FOR Si\\LEM.GENERATING STATION, UNII NO. 1 1.-)
~Je have reviewed the inforrnatiot~ provided by your letters dated April 25, May 11,:
June 1, July 13, and August 14, 1979 in response t~ IE Bulletins 79-0GA and l9~06A~ Revision 1, for tha Salem Generating Station~ Unit No. l. The enclosure.
provides our evaluati_on of _your responses with respect to their specifkity~
completeness, and responsiveness to the intent *of said bulletins. Jn this regard*
we have found that you have taken appropriate actions to.meet the requirements of IE Bul]etins 79-0GA and 79~06A, Revision 1~
It should be noted tb*t the staff review of the Three Mile I~land; Unit 2 acciden is continuing. Consequently, other corrective actipns may be requfred at a 'iater' date.
For example~ IE Bulletii1 79-06C i'Jas* issued on July 26~ 19'79 requiring new consjd~rations for-operation qf,the reattor coolant pumps follpwing an accident~ 1 Our review of the ~Jestin9house Operating Plants Owners' Group response to Items 2.
and 3 of Bulletin ::_9-06C (Westinghouse 1"cworts* l*ICAP-95S4 and l<JCAP-9600, *respettiv and your response dated August 29~ 1979 is continuing.
Yo~ will be ihformed'of the resuli::s of this review by separate correspondence.
In.additiOn, new require-:
ments may result from our generic-rev*im'I* of prpcedµres for operatfog ~Jestinghouse designed plants, our review of plant perfQrmancc during feedwater *transients: and*
sma 11-break loss-of-coo 1 ant acc*i dents~ C}nd* _from our review of l i ce1isees 1
. response tq the requirements delineated fo NUREG..:.0578.
~ *
- Sincerely~
Original Signed By
Enclosure:
Evaluation of Licensee's Responses A.
I
- ~
. MlC Jl.!'Oru\\I 318 (9*76) Nr<lCM O:ZMll U.iJ. QQVts:RNMl'!INT PRINTING OPFICE: ~978... 20!5 - l'EH).
UNITED ST ATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 Mr. F. P. Librizzi, General Manager Electric Production Public Service Electric & Gas Company 80 Park Place, Room 7221 Newark, New Jersey 07101
Dear Mr. Librizzi:
December 31, l 979
SUBJECT:
NRC STAFF EVALUATION OF PSE&G RESPONSES TO IE BULLETINS79-06A AND 79-06A, REVISION l, FOR SALEM GENERATING STATION, UNIT NO. l We have reviewed the information provided by your letters dated April 25, ~:ay 11, June l, July 13, and August 14, 1979 in response to IE Bulletins79-06A and 79-06A, Revision 1, for the Salem Generating Station, Unit No. 1. The enclosure provides our evaluation of your responses with respect to their specificity, completeness, and responsiveness to the in~ent of said bulletins.
In this regard, we have found that you have taken appropriate actions to meet the requirements of IE Bulletins79-06A and 79-06A, Revision 1.
It should be noted that the staff review of the Three Mile Island, Unit 2 accident is continuing. Consequently, other corrective actions may be required at a later date.
For example, IE Bulletin 79-06C was issued on July 26, 1979 requiring new considerations for operation of the reactor coolant pumps following an accident.
Our review of the Westinghouse Operating Plants Owners 1 Group response to* Items 2 and 3 of Bul.letin 79-06C (Westinghouse reports WCAP-9584 and WCAP-9600, respectively) and your response dated August 29, 1979 is continuing.
You will be. informed of the resuJts of this review by separate correspondence.
In addition, new require-ments may result from our generic review of procedures for operating Westinghouse-designed plants, our review of plant performance during feedwater transients and small-break loss-of-coolant accidents, and from our review of licensees' responses to the requirements delineated in NUREG-0578.
Since~
0.: S~:Wencer, Chi~-'L.--
Enclosure:
Evaluation of Licensee's Responses to IE Bulletins79-06A and 79-06A, Revision 1 Operating Reactors Branch #1 Division of Operating Reactors
Mr. F. P. Librizzi Public Service Electric and Gas Company cc: Mark J. Wetterhahn, Esquire Conner, Moore and Corber Suite 1050 1747 Pennsylvania Avenue, NW Washington, D. C.
20006 Richard Fryling, Jr., Esquire Assistant General Solicitor Public Service Electric and Gas Company 80 Park Place Newark, New Jersey 07101 Gene Fisher, Bureau of Chief Bureau of Radiation Protection 380 Scotch Road Trenton, New Jersey 08628 Mr. Hank Midura, Manager Salem Nuclear Generating Station Public. Service Electric and Gas Company 80 Park Place Newark, New Jersey 07101 Mr. R. L. Mittl, General Manager Licensing and Environment Public Service Electric and Gas Company 80 Park Place
.Newark, New Jersey 07101 Salem Free Library 112 West Broadway Salem, New Jersey 08079 Leif J. Norrholm U *. S. Nuclear Regulatory Cammi ssion Drawer I Hancocks Bridge,. New Jersey 08038 December 31, 1979
INTRODUCTION EVALUATION OF LICENSEE'S RESPONSES TO IE BULLETINS79-06A AND 79-06A (REVISION l)
SALEM GENERATING STATION, UNIT l~O. l - DOCKET NO. 5U-L72 By letters dated April 14, 1979 and April 18, 1979, we transmitted I&E Bulletins No.79-06A and No.79-06A (Revision l), respectively, to Public Service Electric and Gas Company (PSE&G or the licensee).
These Bulletins specified actions to be taken by the licensee to avoid occurrence of an event similar to that which occurred at Three Mile Island, Unit.No.~~TMI-2), on March 28, 1979.
By letters dated April 25 and June l, 1979, PSE&G provided their response in conformance with the requ1rements of these Bulletins for the Salem Generating Station, Unit No.
- l.
PSE&G supplemented these responses by letters dated July 13 and August 14, 1979, providing clarification and elaboration of certain of the Bulletin Action Items in response to our expressed concerns.
Our evaluation of the responses, as supplemented, is given below.
Evaluation In this evaluation, the paragraph numbers correspond to the bulletin action items and to the licensee's response to each action item.
- l.
On April 20, 1979, an NRC briefing team provided a detailed review of the circumstances described in Enclosure l of IE ~ulletin 79-05 and the preliminary chronology of the TMI-2 accident Gncluded in Enclosure l of IE Bulletin 79-05A)to licensed station personnel and plant management.
2 -
The briefing ~earn consisted of an Office of Inspection and Enforcement (IE) Group Leader, an Operator Licensing Branch (NRR/OLB) representative, and the facility Principal/Resident Inspector.
Attendance was documented, with any missing personnel ~eing briefed at a later date by the NRC Principal/Resident Inspector.
The NRC briefing also provided a detailed review of Action Item Nos. l.a and l:b-of-IE ~ulletin 79-06A.
In their response, PSE&G stated that an overall package of TMI-related training will *include additional review of the sequence cif events at TMI-2 and additional procedural requirements regarding the termination of engineered safety features.
As part of PSE&G 1 s existing operator qualification program, documentation is maintained of lecture attendance and procedure review.
We consider these actions to be acceptable responses to Action Item No. l.
- 2.
Action Item 2 of the Bulletin requested licensees to review actions required by operating procedures for coping with transients and accidents, with particular attention to (a) recognition of the possibility of forming voids large enough to compromise core cooling capability, (b) action required to prevent the formation of such voids, and (c) action required to enhance core cooling in the event such voids are formed.
Emphasis in (a) was placed on natural circulation capability.
In their response to this Bulletin Action Item, PSE&G referenced the work of the Westinghouse Operating Plants Owners 1 Group (PSE&G is a member of this Owners 1 Group). In conjunction with Westinghouse, the Owners 1 Group has developed generic guidelines for emergency operating procedures regarding srnal 1-t>reak loss-of-coolant accidents tLUCAs).
In its November 5, and December B, 1979 letters to the Owners' Group, the staff approved these guidel ir1es for implementation by licensees with Westinghouse-designed reactors.
The Owners* Group and Westinghouse have also developed generic guidelines for emergency procedures regarding natural circulation.
These generic guide-1 ines were submitted as part of the Owners* Group response to the requirements of NUREG-0578 regarding inadequate core coo1ing.
PSE&G has committed to incorporate the generic guidelines developed by the Owners* Group into its plant procedures and operator training program.
In order to satisfy NUREG-0578 requirements, this effort should b.e complete by January 1980.
We wi 11 verify that the guidelines have been properly implemented.
Procedures based on these generic guidelines represent an acceptable method of complying with Bulletin Action Item No. 2.
PSE&G has also installed a computer program which provides the operator additional information relative to recognizing the possible formation of voids in the primary coolant system.
This program computes the margin to saturation conditions based on the hottest in-core thermocouple reading and the reactor coolant system pressure.
This program indicates the degrees of subcooling.
An alarm is generated if 50° of subcooling does not exist whenever reactor power is less than 0.25%.
An alarm is also generated if the difference between actual and saturation pressure is less than 200 psi.
4 -
We find that licensee has provided an acceptable response to Bulletin Action Item No. 2.
- 3.
The pressurizer low-level bistables for safety injection are in a tripped condition.
They will be maintained in this condition until the design change to a revised low pressure logic is completed.
This design change moves the level input requirement and changes the pressure coincidence to a two-out-of-three logic for initiation of safety injection.
Existing procedures direct the operators to manually initiate any protec-tion functions, if the automatic initiation fails.
Although this ensures manual initiation of safety injection on low pressurizer pressure, addi-tional training was given to operating personnel in light of the TMI-2 accident which addressed the revised logic.
This training effort was completed in August 1979.
We find the licensee 1 s response to Bulletin Action Item No. 3 acceptable.
- 4.
The Salem Unit No. l design provides for automatic initiation of containmert isolati'on upon safety injection actuation, as called for in the bulletin.
This aspect of the licensee 1 s response is therefore acceptable.
Containment isolation consists of a Phase A and a Phase B isolation.
Phase A.involves closure of automatic valves in all nonessential process lines; Phase B isolates all remaining process lines, except for safety injection, containment spray, and auxiliary feedwater.
The reactor coolant pump seal water discharge line is isolated upon a Phase A signal'.
The seal water supply line is not provided with isolation valves.
The component cooling water supply and return lines for the reactor coolant pumps are isolated by a Phase B signal.
The reactor coolant pumps do not trip automatically on either isolation signal.
Therefore, the pumps must be manually tripped following a Phase B isola-tion, since component cooling to the motor coolers and thermal barriers is lost.
We find that the licensee has adequately addressed the concerns expressed in Bulletin Action Item No. 4.
- 5.
The auxiliary feedwater system is automatically initiated at Salem Unit No. 1, with no operator action required in order to ensure adequate flow.
Therefore, Bulletin Action Item No. 5 does not apply to this plant.
- 6.
Current Salem Unit No. 1 procedures assure that operating personnel are aware of plant indications available to detect an open pressurizer PORV.
These procedures include instructions to isolate the PORV if it is stuck open.
In their response to this item, PSE&G also identified the information that is available to the operator which provides indication of an open PORV.
We find the licensee 1 s response to Bulletin Action Item No. 6 acceptable.
/a.
In its July 13, 1979 supplemental response to this item, PSE&G stated that a complete review of the Salem Unit No. l station procedures indicated that the only engineered safety feature which is overridden is safety injection.
PSE&G referenced the works of the Westinghouse Operating Plants Owners 1
Group concerning resolution with the NRC staff of the conditions under which safety injection may be overridden and terminated.
The PSE&G response included a commitment to incorporate the resolution of this issue between the Owners 1
Group and the staff into the station procedures.
PSE&G also stated that it had discovered that is was possible to'inadvert-ently override the RMS interlock on the Containment Ventilation System by improper operation of the reset functions.
To prevent occurrence of this situation, addition~l instructions were issued to the operators and were included in the procedures and the -Operator training program.
Because of the discovery of this problem, PSE&G undertook an investigation to verify that there were no similar situations.
The results of that review verified that safety functions are not overidden and are allow~d to go to completion, as considered in the plant design bases.
We find that the licensee has addressed the concerns expressed in this Bulletin Action Item in an acceptable manner.
lb.
As.stated above, PSE&G committed to the resolution of the issue regarding termination of safety injection between the Owners 1 Group and the staff.
In our November 5, 1979 letter to the Owners 1 Group, we approved generic guidelines for emergency procedures regarding small break LOCAs for incorporation by licensees into their plant procedures.
These approved guidelines include the following criteria for termination of safety injection:
(1)
The reactor coolant system pressure is greater than 2000 psig and increasing, and (2)
The pressurizer water level is greater than 50% of span, and (3)
The reactor coolant indicated subcooling is greater than (insert plant-specific value of subcooling based on full power normal operation), and (4)
The water level in at least one steam generator is in the narrow range span, or in the wide range span at a level sufficient to assure that the u-tubes are covered.
Details of our evaluation of this issue will be included in the forth-coming staff report (NUREG-0611) of our generic review of Westinghouse-designed operating plants.
We will verify that the approved Westinghouse generic safety injection termination criteria have been properly incorporated in the Salem Unit plant procedures.
Pending such verification, we find that the licensee 1 s response to this Bulletin Action Item is acceptable.
7c.
In their April 25, 1979 response to this item, PSE&G stated that Westing-house had advised it to manually trip all reactor coolant pumps in LOCA 1 s and steam line break accidents when the following conditions were satisfied:
verification of ECCS operability, decreasing reactor colant system pressure, or occurrence of Phase B containment isolation.
- Following discussion with the staff about the April 2o response, PSE&G, in its July 13, 1979 letter, committed to the resolution of this issue between the staff and the Owners' Group.
On July 26, 1979 IE Bulletin 79-06C superseded Action Item 7.c of Bulletin 79-06A.
Bulletin 79-06C required that, as a short-term action, licensees were to trip all reactor coolant pumps after an initia.
Jf safety injection caused by low reactor coolant system pressure.
In its August 29, 1979 response, PSE&G stated its conformance with this require-ment.
This action was to remain in effect until the results of analyses defined in IE Bulletin 79-06C had been used to develop new guidelines for operator action.
We have completed our review of the reactor coolant pump trip issue with the Owners' Group.
The generic guidelines for emergency procedures regarding small break LOCAs which we approved in our November 5, 1979 letter, contain the approved pump trip criteria for Westinghouse-designed operating plants.
Basically they are as follows:
Stop all reactor coolant pumps after high pressure safety injection pump operation has been verified and when the wide range reactor coolant pressure is at (plant-specific pressure derived from secondary system relief capacity, primary to secondary system pressure difference, and instrument inaccuracies).
9 -
The details of our review of the pump trip issue are reported in the forthcoming NUREG-0623.
Since the licensee has committed to incorporate the pump trip criteria as specified in the approved generic guidelines into the Salem Unit l proce-dures, we find the licensee 1 s response to this Bulletin Action Item acceptable.
7d.
In its response to this item, PSE&G stated that a portion of the TMI-related training to be accomplished in August 1979 would instruct operating personnel not to rely upon a single parameter alone.
PSE&G further stated that the generic guidelines for emergency procedures being developed by the Owners' Group and Westinghouse would include the appropriate additional parameters to be used by the operators for evaluating plant conditions.
PSE&G committed to incorporate the generic guidelines into the Salem Unitl'Jo. 1 procedures* after they have been approved by the staff.
Pending our verification of the licensee 1 s commitment to incorporate the approved guidelines into the plant procedures, we find the licensee 1 s response to this Bulletin Action Item acceptable.
- 8.
This Bulletin Action Item required the review of alignment and alignment requirements and controls for all safety-related valves necessary for proper operation of engineered safety features.
PSE&G has completed the required review and incorporated all necessary changes into the plant procedures.
The status of key safety system valves was verified by visual examination shortly after the TMI-2 accident.
-1
~
I
~
I
- All safety-related valves which are locked in the proper position are verified by surveillance procedure.
Valve positions which are changed from normal positions are recorded in the valve deviation log and the operator 1 s shift log.
All system valve lineups were completed prior to plant startup.
We find the licensee 1 s response to Bulletin Action Item No. 8 acceptable.
- 9.
In Bulletin Action Item No. 9, licensees were requested to review their procedures to assure that radioactivity will not be inadvertently released from containment.
Particular emphasis was placed on resetting of engineered safety features CESFs and the effect of this action on valves controlling the release of radioactivity.
In its response, PSE&G identified all systems which are designed to transfer potentially radioactive fluids from containment.
For each of these systems, PSE&G addressed high radiation interlocks, containment isolation (Phase A and Phase B), and operability assurances, as requested.
Two instances were identified, the Reactor Coolant Drain Tank pump dis-charge line and the Pressurizer Relief Tank gas analyzer line, which could result in the inadvertent transfer of radioactive material from the containment.
PSE&G stated that design changes to revise the control circuitry to prevent the occurrence of an open pathway in these two instances would be implemented before plant startup for Cycle 2.
(At the time of their response, Salem Unit No. l was shut down for refuelinq'.
i We find the licensee 1 s actions in response to Bulletin Action Item No. 9 acceptable.
- 10.
For safety-related systems, Action Item 10 required that licensees review and modify, as necessary, maintenance and test procedures to ensure that they require that:
(a) redundant systems are operable before a system is taken out of service, (b) systems are operable when returned to service, and (c) operators are made aware of the status of these systems.
PSE&G has reviewed station procedl!Jres and revised them,where necessary,to detail requirements for verifying the operability of redundant equipment prior to removing safety-related equipment from service and verifying the operability of equipment when it is returned to service.
Both systems level considerations and individual safety-system equipment are addressed.
PSE&G stated that the Shift Supervisor/Senior Shift Supervisor is responsi-ble for approving all requests for removal of equipment for service.
The control operator prepares the necessary administrative tags which are used to identify equipment removed from service.
The equipment operator places these tags on the equipment taken out of service.
The control operator also indicates control room equipment out-of-service by the use of tags and other identification methods.
In order to adequately handle system status at shift change, PSE&G developed and implemented a formal shift turnover procedure.
.~ We find that the licensee has adequately addressed all of the concerns expressed in Bulletin Action Item No. 10.
ll.
Station Supervisory Letter SL-9, 11 Notification of Federal and State
- Agencies, 11 has been revised and issued to require notification of the NRC within one hour of the plant being in an uncontrolled or unexpected condition.
Telephone lines to establish the required open line of communi-cation between the Salem plant and IE Region I via Bethesda, Maryland have been installed and are now functional.
Additional telephone lines to provide communications from the Salem plant to the NRC for radiation protection/chemistry matters will be installed after receipt of orders from NRC.
The Station Emergency Plan has been revised to include the location and use of these lines.
The licensee 1s actions are considered an acceptable response to Bulletin Action Item No. 11.
- 12.
In its response to this i tern, PSE&G state d that it was continuing to review the modes for controlling hydrogen_in the reactor coolant system.
All procedural changes for coolant system and containment hydrogen control were to be implemented* p"ri6r to Unit l 1s return-to power from the recent outage.
The options for removal of hydrogen from the reactor coolant system include (l) stripping hydrogen from the reactor coolant to the pressurizer vapor space and venting to the pressurizer relief tank, (2) removing hydrogen from the reactor coolant system via the letdown line and stripping it in the volume control tank and venting through the waste gas system, and (3) in the event of a LOCA, hydrogen would vent with steam into containment.
PSE&G also described modes and proc~dures for removal of a non-condensible gas bubble from the primary coolant system while maintaining core cooling.
In addition, PSE&G is participating in the Westinghouse Operating Plant Owners' Group efforts to develop general guidelines for emergency opera-tional procedures regarding inadequate core cooling in response to the requirements of NUREG-0578.
Treatment of noncondensible gas in the reactor coolant system is being considered in the development of these guidelines.
During recent discussions with PSE&G, we have been informed that each of the options for dealing with hydrogen described above will be incorporated in the plant procedures where needed to address various plant conditions.
This implementation will be completed by January l, 1980. __ We will verify that this commitment has been fulfilled.
We find that the licensee's actions in response to the concerns expressed in Action Item No. 12 are acceptable.
- 13.
This Bulletin Action Item requested licensees to propose changes to the plant Technical Specifications, as required, which had to be modified as a result of implementing Action Items l through 12.
In their June 1, 1979 letter, PSE&G identified the design changes and Technical Specification changes that were required, up to that time, to implement Bulletin Action Items l through 12.
According to PSE&G, the only required Technical Specification change reflected deletion of the coincident Pressurizer Low Level and Low Pressure Signals for initiating safety injection.
As discussed in our evaluation of Bulletin Action Item No. 3, the revised design consists of a two-out-of-three coincidence of ecessurizer Lo~ Pressure Signals.
We find that the licensee has made an adequate response to Bulletin Action Item No. 13.
CONCLUSIONS Based on our review of the information provided by the licensee, we conclude that the licensee has correctly interpreted IE Bulletins79-06A and 79-06A, Revision l.
The actions taken demonstrate the licensee 1 s understanding of the concerns arising from the Three Mile Island, Unit No. 2 accident in relation to their implications on his own operations, and provide added assurance for the protection of the public health and safety during plant operation.
This conclusion, notwithstanding, should be recognized that further actions may result from the staff 1s ongoing review of operating plants using nuclear steam supply systems designed by Westinghouse.
Additional changes may result from the requirements contained in NUREG-0578 (e.g., the actions being taken for Item 6 of Bulletin 79-06A regarding the PORV 1s). Our evaluation of such matters will be provided in other reports.