ML17250A333
| ML17250A333 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 05/28/1980 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | White L ROCHESTER GAS & ELECTRIC CORP. |
| References | |
| IEB-79-06A, IEB-79-6A, NUDOCS 8006190103 | |
| Download: ML17250A333 (24) | |
Text
,jIKXtIii L';Ir'i.'.'.'I'.i:i d I Lt,'ilorIl(
Docket No. 50-244 8
~98O Nr. Leon D. Mhite, Jr.
Vice-Pres ident-El ectri c and Steam Production Rochester Gas
& Electri c Corporation 89 East Avenue-Rochester, New York 14649
Dear Nr. Mhite:
DISTRIBUTION:
NRC PDR Local PDR TERA N SIC NRR Reading ORB g5 Reading DEisenhut GLai nas DCrutchfiel d JShea SNowi cki HSmi th PDO'ei 1 ly
- EJordan, IE WRussel 1
oIE (3)
OELD, Attorney ACRS ( 16)
SUBJECT:
NRC STAFF EVALUATION OF RG&E RESPONSES TO IE BULLETINS 79-06 AND 79-06A, REVISION I, FOR ROBERT E.. GINNA NUCLEAR POWER PLANT, UNIT 1 1
Me have reviewed the information provided by your. letters dated April 28 and June 22, 1979, in response to IE Bul 'leti ns79-06A and 79-06A, Revision 1, for Rober t E. Ginna Poorer Plant, Unit 1.
The enclosure provides our eval uation of your responses with respect to their specificity, completeness, and responsiveness to the bul 1etins.
In this regard, we have found that you have taken appropl late actions to meet the requirements of IE Bul 1eti ns79-06A and 79-06A, Revisi on 1 ~
It should be noted that the staff review of the Three Nile Island,, Unit 2 accident is continuing.
Consequently, other corrective actions may be required at a
'1ater date.
For exampl e, IE Bul 1etin 79-06C was issued on July 26. 1979, requiring new considerations for operation of the reactor coolant pumps fol 1 owing an accident.
Our reviews of the Mestinghouse Owners Group response to Items 2 and 3 of Bu'lletin 79-06C (Mestinghouse reports MCAP-9584 and MCAP-9600, respectively) are documented in NUREG-0623 and NUREG-061 1, respectively.
You wi 1 1 be kept informed regarding the require-ments f'r the Ginna Nuclear Power Plant resul ting from these reviews by separ ate correspondence.
Sincerely, dri8i~al signed b r emnnis M. Crutchfi el'ennis f1. Crutchfield, Chief Operating Projects Branch 5
Division of Licensing
Enclosure:
Eval uation of Licensee '
Responses to IE Bul1etins79-06A and 79-06A, Revision 1
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Docket No. 50-244 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 May 28, 1980 Mr. Leon D. White, Jr.
Vice-President-Electric and Steam Production Rochester Gas 8 Electric Corporation 89 East Avenue Rochester, New York 14649
Dear Mr. White:
SUBJECT:
NRC STAFF EVALUATION OF RG&E RESPONSES TO IE BULLETINS 79-06 AND 79-06A, REVISION 1, FOR ROBERT E.
GINNA NUCLEAR POWER PLANT, UNIT 1 We have reviewed the information provided by your letters dated April 28 and June 22, 1979, in response to IE Bulletins79-06A and 79-06A, Revision 1, for Robert E. Ginna Power Plant, Unit 1.
The enclosure provides our evaluation of your responses with respect to their specificity,-completeness, and responsiveness to the 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 reviews of the Westinghouse Owners Group response to Items 2 and 3 of Bulletin 79-06C (Westinghouse reports WCAP-9584 and WCAP-9600, respectively) are documented in NUREG-0623 and NUREG-0611, respectively.
You will be kept informed regarding the require-ments for the Ginna Nuclear Power Plant resulting from these reviews by separate correspondence.
Sincerely, Dennis M. Crutchfield, Chic Operating.Projects Branch 5
Division of Licensing
Enclosure:
Evaluation of Licensee's Responses to IE Bulletins79-06A and ?9-06A, Revision 1
I
'I Mr.'"Leon D. White,'Jr.
~II~
ro I
cc w/enclosure:
Harry H. Voigt, Esquire
- LeBoeuf, Lamb, Leiby 8 MacRae 1757 N Street, N.
W.
Washington, D. C.
20036 Mr. Michael Slade 12 Trai lwood Circle Rochester, New York 14618 Rochester Coamittee for Scientific Information Robert E. Lee, Ph.D.
P. 0.
Box 5236 River Campus
,Station Rochester, New York 14627 2
N May '28, 1980 Director, Technical Assessment Division Office of Radiation Programs (AW-459)
U. S. Environmental Protection Agency Crystal Mall 0'2 Arlington, Virginia 20460 U. S.
E nvi ronment a 1 P rotect i on Agency Region II Office ATTN:(
E IS COORDINATOR 26 Federal Plaza New York, New York 10007 Jeffrey Cohen New York State Energy Office Swan Street Building Core 1, Second Floor Empire State Plaza
- AIbarp, New York 12223 Director, Technical Development Programs State of New York E nergy 0ffice Agency Building 2 Empire State Plaza
Supervisor of the Town of Ontario 107 Ridge Road West
- Ontario, New York 14519 Mi. Husain
,I, Yankee Atomic Electric Company 20 Turnpike Road
- Westboro, Massachusetts 01581 Robert W. Jurgensen American Electric Power Sei vice
'Corporation 2 Broadway New York, New York 10004
Herbert Grossman, Esq.,'hai rman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Coranission Washington, D., C.
20555 Dr. Richard F.
Cole'tomi c Sa fety and Licens i ng Board U. S. Nuclear Regulatory Comission Washington, D. C.
20555 Dr.
Emmeth A'uebke Atomic Safety and Licensing Board U. S.~Nuclear Regulatory Corrrnission'ashington, D. C.
20555 Mr. Thomas B. Cochran Natural Resources Defense Council, Inc.
1725 I Street,',N.
W.
Suite 600 Washington, D.,C.
20006 Rik W. Wells Northeast Utilities P. 0.
Box 270 Hartford, Connecticut 06101 r>>
W. J. Cahill Consolidated Edison 4 Irving Place New York, New ~York 10003
l 1
Mr. Leon D. White, Jr.
3 May 28, 1980 cc w/enclosure:
Peter W. Lyon Power Authority of the State of new York 10 Columbus Circle New York, New York 10019 lohn A. Ahladas Virginia Electric 5 Power Company One James River Plaza P. 0.
Box 26666 Richmond, Virginia 23261 Frank P. Librizzi Public Service Electric 5 Gas Co.
80 Park Place
- Newark, New Jersey 07101 Dave Waters Carolina Power 5 Light P. 0.
Box 1551
- Raleigh, North Carolina 27602 J. J.
Carey Duquesne Light P.
0.
Box 1551 Shippingport, Pennsylvania 15077 Robert Mecredy Rochester Gas E Electric 89 East Avenue Rochester, New York 14649 Alabama Power Company 600 North 18th Street P. 0.
Box 2641 Birmingham, Alabama 35291 George Liebler Florida Power 5 Light Co.
P. 0.
Box 529100 Miami, Florida 33152 Mark Marchi Wisconsin Public Service Corp.
P. 0.
Box 1200 Green Bay, Wisconsin 54305 F. Pat Tierney, Jr.
Northern States Power Company Route 2
Welch, Minnesota 55089 Roger Newton Wisconsin Electric Power Company 231 West Michigan Milwaukee, Wisconsin 53201 Cordell Reed Commonwealth Edison Company P. 0.
Box 767 One First National Plaza Chicago, Illinois 60690 Dani el'. Herborn Portland General Electric Company 121 West Salmon Street
- Portland, Oregon 97204 Jerry G. Haynes Southern California. Edison Company 2244 Walnut Grove Avenue
- Rosemead, California 91770 Jim Gormly Pacific Gas 5 Electric Company 77 Beale Street San Francisco, Califor'nia 94106 Bill Layman Electric Power Research Institute 3412 Hillview Avenue Palo Alto, California 94303 Bob Szalay Atomic Industria]
Forum 7101 Wisconsin Avenue.
77 Grove Street
- Rutland, Vermont 05701
I
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EVALUATION OF LICENSEE'S RESPONSES TO IE BULLETINS79-06A AND 79-06A (REVISION 1)
ROBERT E.
GINNA NUCLEAR POWER PLANT, UNIT 1
DOCKET NO. 50-244 INTRODUCTION By letters dated April 14, and April 18,
- 1979, we transmitted IE Bulletins 79 06<
and 79-06A (Revision 1), respectively, to Rochester Gas and Electric Corpora-tion (RG8E or the licensee).
These bulletins specified actions to be taken by the licensee to avoid occurrence of an event similar to that which occurred on March 28, 1979 at Three Mile Island, Unit No.
2 (TMI-2).
By letter dated April 28,
- 1979, the licensee provided its response to the aforementioned bulletins for Robert E.
Ginna Nuclear Power Plant Unit 1 (Ginna).
The licensee supplemented its response by letter dated June 22, 1979, providing clarification and elaboration of certain of the Bulletin Action Items in response to our expressed concerns.
Our evaluation of the licensee's responses, as supplemented, is provided 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.
In Bulletin Action Item No.
1, licensees were requested to review the description of ci rcumstances described in Enclosure 1 of IE Bulletin 79-05 (issued to all licensees with Babcock 8 Wilcox (B8W)-designed plantsfor
- action, and to all other licensees for information) and the preliminary
, chronology of the TMI-2 accident included in Enclosure 1 to IE Bulletin 79-05A (same distribution as IE Bulletin 79-05).
(a)
This review should be directed toward understanding:
(1) the ff extreme seriousness and consequences of the simultaneous blocking of both auxiliary feedwater trains at the TMI-2 plant and other actions taken during the early phases of the accident, (2) the apparent
~,
operational errors which led to the eventual core
- damage, (3) that the potential exists, under certain accident or transient condi-tions, to have a water level in the pressurizer simultaneously with the reactor vessel not full of water, and (4) the necessity to systematically.analyze plant conditions and parameters and take appropriate corrective a'ction.
~ <<C (b)
Operational personnel should be instr'ucted to:,
(1) not override automatic action of engineered safety features unless continued operation of engineered safety featureswill result in unsafe plant conditions (see Section 7a.);
and (2) not make operational decisionq based solely on a single plant parameter indication when one or more confirmatory indications are available.
(c)
All licensed operators and plant management and supervisors with operational responsibilities were to 'participate in this review and such participation was to be documented in plant records.
On April 18,
- 1979, an NRC briefing team provided a detailed review of the circumstances described in Enclosure 1 of IE Bulletin 79-05 and the preliminary chronology of the TMI-2 accident included in Enclosure 1 of IE Bulletin 79-05A to a majority of the licensed operators and plant management.
The briefing team consisted of an IE Section
- Leader, an Operator Licensing Branch (OLB/NRR) representative,.
and the'facility Principal/Resident Inspector.
Attendance was documented with any absentees being briefed at a later date by the NRC Principal/Resident Inspector.
The NRC briefing also provided a detailed review of Items 1. a and 1.b of IE Bulletin 79-06A.
We consider the NRC briefing to be an acceptable response to Bulletin Action Item No. 1..
2.
Action Item No.
2 of the Bulletin requested licensees to review actions required by operating procedures for coping with transients and acci-dents, with particular attention to (a) recogriition of the possibility for 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 response to this item, the licensee developed emergency procedures to recognize and deal with the formation of voids.
Procedure 0-8, "Natural Circulation," provides instructions on identifying whether or not heat is being removed from the primary system by natural circulation.
Proce-dure E-l.5, "Void Formation in the RCS," provides guidance on how to eliminate voids from the RCS.
Guidance has also been provided on removing heat from the RCS if neither forced nor natural circulation is available, and if the break flow from a LOCA is not sufficient to remove decay heat.
Another new procedure, E-1.6, "CVCS Breaks,"
was developed which provides for better operator response to breaks in the CVC system.
In addition to the above, the licensee modified the following emergency procedures to incorporate actions to prevent inadvertent formation of voids:
E-l. 1, "Safety Injection Initiation," E-1'.2, "Loss of Coolant Accident," E-l.3, "Steam Line Break Accident," and E-1.4, "Steam Generator Tube Rupture Accident."
The licensee identified the information then available in the control room to assist the operator in recognizing conditions favorable for void formation.
These include saturation conditions correlations and core exit thermocouple indications.
Among the long-term efforts pursued by the licensee in reponse to this i'tern are additional instrumentation and displays available to the operator.
In addition, the licensee participated, as a member of the Westinghouse Owners Group in the effort to develop generic guidelines for emergency procedures.
In our November 5 and December 6,
1979 letters to the Owners
- Group, we approved the Westinghouse generic guidelines regarding small break LOCAs for implementation by licensees with Westinghouse-designed reactors.
The Owners Group, in conjunction with Westinghouse, has also developed generic guidelines for emergency procedures regarding natural circulation.
These generic guidelines were submitted on December 28,
- 1979, as part of the Owners Group response to the requirements of
Item 2.1.9 of NUREG-0578 regarding inadequate core cooling.
In order to satisfy NUREG-0578 requirements, the licensee should have incorporated the guidelines into the Ginna procedures (small break LOCA guidelines by January 1,
1980 and'nadequate core cooling guidelines by January 31, 1980).
The Office 'of Inspection and Enforcement will verify that acceptable guidelines have been properly implemented.
Procedures based on these generic guidelines represent an acceptable method of complying, with Bulletin Action Item No.
2..
Me find that the licensee has provided an acceptable response to Bul,letin Action Item No.
2.
Bulletin Action Item No.
3 requested that licensees with facilities that used pressurizer water level coincident with pressurizer pressure for automatic initiation of safety injection into the reactor coolant system trip the low pressurizer
'level setpoint bistables such that, when the pressurizer pressure reached the low setpoint, safety injection would be initiated regardless of the pressurizer level.
The pressurizer level bistables could be returned to their normal operating positions during the pressurizer pressure channel functional surveillance tests.
In its initial response to this bulletin action item, the licensee tripped the low pressurizer level bistables and instructed operations personnel to manually initiate safety injection whenever the pressurizer pressure reached the low pressure setpoi nt.
I II On June 7, 1979, the licensee submitted a proposed modification to the safety injection actuation logic and appropriate ch'anges to the plant Technical Specifications for staff review.
The design change requested was that the safety injection signal be developed by a two-out-of-three logic based on low pressurizer pressure
- alone, instead of a one-of-three logic based on low pressurizer pressure coincident with low pressurizer level.
We reviewed and approved the licensee's
- proposal, issuing Amendment No.
27 toithe Ginna operating license on June 15, 1979.
Based on our review, we find the licensee's response to Bulletin Action Item No.
3 acceptable.
4.
Bulletin Action Item No.
4 requested that licensees review the containment isolation initiation design and procedures, and implement all changes necessary to permit containment isolation, whether manual or automatic, of all lines whose isolation would not degrade needed safety features or cooling capability, upon automatic initiation of safety injection.
The Ginna design provides for automatic initiation of containment isola-tion upon safety injection actuation, as called for in the bulletin.
This aspect of the licensee's response is therefore acceptable.
In its June 22, 1979 supplemental
- response, the licensee identified all valves which are closed on either a containment isolation signal or a containment ventilation isolation signal.
The remaining lines, which are not isolated on a containment isolation or containment ventilation isolation signal, are those required for safety functions.
In addition, component cooling water and seal injection lines to the reactor coolant pumps are not isolated.
Therefore, the reactor coolant pumps remain operable under this isolated condition.
The licensee completed a review of safeguard logic schemes and concluded that no changes were required with respect to containment isolation.
The licensee also verified that all lines which penetrate "containment that are not required for safety features or reactor coolant pumps are isolated by locked valves, normally closed valves, or automatic valves that close on an isolation signal.
Based on our review of the licensee's
- response, we find that the licensee has adequately addressed the concerns expressed in Bulletin Action Item No. 4.
5.
In Bulletin Action Item No. 5, licensees with facilities at which the auxiliary feedwater system is not automatically initiated were requested
to prepare and implement immediately procedures which required the stationing of an individual (with no other assigned concurrent duties and in direct and continuous communication with the control room) to promptly initiate adequate auxiliary feedwater to the steam generator(s) for 'those.
transients or accidents, the consequences of which could be limited by such action.
I The auxiliary feedwater system at Ginna is automatically initiated, with no operator action required in order to ensure adeqUate flow.
Therefore, Bulletin Action Item No.
5 does not apply to this plant.
6.
Bulletin Action Item No.
6 requested that licensees prepare'nd implement immediately procedures which:
(a)
Identified those plant indications (such as valve discharge piping temperature, valve position indication, or valve discharge relief tank temperature or pressure indication) whicg plant operators could utilize to determine that the pressure power operated 'relief valve(s) are open, and t
(b)
Directed the plant operators to manually close the power-operated relief block valve(s) if the reactor coolant system pressure had been reduced to below the set point for normal, automatic closure of the power-operated relief valve(s) and the valve(s) remained stuck open.
Emergency Procedure E-15.1, "Malfunction of Pressurizer Power Relief or Safety Valves," lists the indications which plant operators may use to determine that the, pressurizer power-operated relief valve(s) are open.
On April 23, 1979, the licensee revised this procedure to implement closure of the block valve(s) if the PORV(s)'is stuck-open.
Based on our
- review, we find the licensee's response to Bulletin Action Item. No.
6 acceptable.
7.
In Bulletin Action Item No. 7, licensees were requested to review the action directed by the operating procedures and training instructions to ensure-that:
(a)
Operators do not override automatic actions of engineered safety
- features, unless continued operation of engineered safety features
\\
would result in unsafe plant conditions.
For example, if continued operation of engineered safety features would threaten reactor vessel integrity, then the HPI system should be secured (as noted in b(2) below).
(b)
Operating procedures currently, or are revised to, specify that, if the high pressure injection (HPI) system had been automatically actuated because of a low pressure condition, it must remain in 9
operation until either:
N (1)
Both low pressure injection (LPI) pumps are in operation and flowing for 20 minutes or longer, at a rate which would assure stable plant behavior, or (2)
The HPI system has been in operation for 20 minutes, and all hot and cold leg temperatures are at least 50 degrees Fahren-heit below the saturation temperature for the existing RCS pressure.
If 50 degrees subcooling cannot be maintained after HPI cutoff, the HPI shall be reactivated.
The degree of subcooling beyond 50 degrees and the length of time HPI is in operation shall be limited by the pressure/temperature considerations for the vessel integrity.
(c)
Operating procedures currently, or are revised to, specify that, in the event of HPI initiation with reactor coolant pumps (RCPs) operating, at least one RCP shall remain operating for two-loop plants and at least two RCPs shall remain operating for 3 or 4 loop
- plants, as long as the pump(s) is providing forced flow.
(d)
Operators are provided additional information and instructions to I
not rely upon pressurizer level indication albne, but to also.
examine'ressurizer pressure and, other plant 'parameter indications in evaluating plant conditions, e.g., water inventory in the reactor, primary system.
In response to Bulletin Action Item No. 7.a, the licensee issued Ginna Station Procedures Change Notice No. 79-1168 to prohibit overriding engineered safety features unless it is determined that continued oper ation of engineered safety features would result in unsafe condi-tions.
This constitutes an acceptable response to Bulleting Action Item No. 7.a.
In response to Bulletin Action Item No. 7.b, the licensee participated in the effort by the Westinghouse Owners Group, in conjunction with Westinghouse, to develop generic guidelines for emergency procedures.
In our November 5 and December 6, 1979 letters to the 'Owners
- Group, we approved generic guidelines for emergency procedures regarding small break LOCAs for implementation by licensees With Westinghouse-designed operating plants.
These approved guidelines include the following criteria (taken from the enclosure to our letter of December 27, 1979) 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 the programmed no-load water level, and (3)
The reactor coolant indicated subcooling is greater than (insert plant-specific value, which is the sum of the errors for the temperature measurement system used and the pressure measurement system translated into temperature using the saturation tables),
and (4)
The water level in at least one steam generator is stable and increasing, as verified by auxiliary feedwatei flow to that unit.
Auxiliary feedwater flow to,the unaffected steam generator should be greater than a value in m sufficient to remove deca heat after 20 minutes followin reactor tri ) until the indicated level is returned to,within the narrow range level instrument.
I V
Details of our evaluation of this issue are included in the report (NUREG-0611) of our generic review of Westinghouse-designed operating plants.
The Office of Inspection and Enforcement will verify that the approved Westinghouse generic safety injection termination criteria have been properly incorporated in the Ginna plant procedures.
Pending such verification, we find that the licensee's actions with regard to this bulletin action item are acceptable.
Another issue on which the Westinghouse Owners Group worked, in conjunction with Westinghouse, to achieve resolution with the staff was the matter of reactor coolant pump operation following a small break LOCA (Bulletin Action Item No.
- 7. c).
On July 26,
- 1979, IE Bulletin 79-06C
'superseded Action Item No. 7.c of Bulletin 79-06A.
Bulletin 79-06C requi red that, as a short-term action, licensees were to trip all reactor coolant pumps after an initiation of safety injection caused by low reactor coolant system pressure.
In its August 29, 1979 response to Bulletin 79-06C, the licensee stated its conformance with this require-ment.
This action was to remain in effect until the results of analyses specified in 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 and December 6,
1979 letters to the Owners Group, contain the approved pump trip criteria for Westinghouse-designed operating plants.
Basically, they are as follows:
(1)
Stop all reactor coolant pumps after high pressure safety injection pump operation has been verified, and when the wide range reactor pressure is at (plant-specific pressure derived from secondary system relief capacity, primary-to-secondary system pressure difference, and instrument inaccUracies).
Appropriate cautions have been included in the guidelines regarding.
isolation of component cooling water to the reactor coolant pumps and maintaining peal injection flow to preclude pump damage dUe to inadequate cooling.
The details of our review of the pump trip issue are reported in NUREG-0623.
Pending conf'irmation by the Office of Inspection and Enforcement that the licensee has incorporated the pump trip criteria as specified in the approved Westinghouse generic guidelines into the Ginna plant procedures, we find the licensee's response to Bulletin Action 'Item No. 7.c acceptable.
In response to BJlletin Action Item, No.
7.d~ the licensee stated that pressurizer level h'ad been eliminated in the diagnostic scheme for the identification of accident category.
In addition, 'pressurizer level was removed as a symptom from all reactor coolant system accident procedures, except for a CVCS break.
The operator is directed'to other indications for accident indentification.
Examples were given.
The licensee listed other available plant paramters which could be used as indicators of water inventory or accident symptoms.
These parameters have been included in the appropriate procedures.
Based on our review of the licensee's response to BUlletin Action Item No. 7.d, we find th'at it is acceptable.
Bulletin Action Item No.
8 required that licensees review alignment and alignment requirements and controls for all safety-"related valves necessary for proper operation of engineered safety features.
10
In response, the licensee reviewed all Periodic Test Procedures concerning the engineered safety features
- systems, and updated
- them, where necessary, to ensure proper valve realignment following testing.
Each procedure directs test personnel to properly realign systems.
Additional steps were added to assure the systems have been realigned for operation.
This additional verification of realignment of systems is performed by personnel other than testing personnel.
The Periodic Test procedures are performed on a scheduled basis according to Technical Specification requirements, and before and after maintenance of engi-neered safety features systems.
Those PT Procedures which were revised were identified and the changes described.
The licensee has developed and incorporated in plant operations new procedures to assure proper system valve lineup.
These system valve position verification procedures are performed on a regularly scheduled basis.
The performance of these procedures is in addition to the valve verification steps included in the Periodic Test procedures.
The Ginna plant Technical Specifications contain no requirements for surveillance of locked valves.
However, surveillance of locked valves is accomplished by valve verification steps in Periodic Tests following System Testing, and by regularly scheduled system valve verification procedures.
Based on our review, we find that the licensee's response to Bulletin Action Item No.
8 is 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 the resetting of engineered safety features (ESFs) and the effects of this action on valves controlling the release of radioactivity.
In its June 22, 1979 supplemental
- response, the licensee listed all lines which are designed to transfer potentially radioactive fluids from containment.
The valves which are closed by a containment isolation 11
signal were listed.
The containment islation signal also trips the containment sump
'pumps and initiates a containment ventilation isolation signal.
The valves closed by a containment venti'lation isolation signal were also identified.
The containment ventilation isolation signal also trips the purge sUpply and exhaust fans.
Resetting of the safety injection signal does not reset containment isolation or containment ventilation isolation and therefore does not reposition any valves.
The licensee identified a number of valves which 'are normally open and would return to the open position upon reset of c'ontainment isolation unless additional actions were taken to preclude this.
Certain of these valves will not reopen if a high radiation condition exists in the steam generator blowdown system.
Current Ginna LOCA and steam line break procedures require that the operators place all containment isolation valve switches in the "closed" position, prior to resetting containment isolation.
The licensee also modified the plant emergency procedures to instruct the operator to place the containment sump "A" pumps in the pull-stop positidn prior to resetting the containment isolation signal.
The licensee also, referenced correspondence between the staff and the licenseeconcerning safety actuation circuits and their overrides.
Besides the above-mentioned considerations',
the licensee also stated that the containment ventilation isolation does not aut'omatically reset when either safety injection or containment isolation is reset.
In order to reset containment, ventilation isolation, the use of a key held by the Shift Foreman is required.
We find that the licensee has adequately addressed'he concerns expressed in Bulletin Action,Item No.
9.
<j The staff's implementation of Item 2.1.4 of NUREG-0578 provides further assurance that the, inadvertent release of radioactivity from containment upon resetting of ESFs will be precluded.
Our review of NUREG-0578 Item 2.1.4 implementation will be reported in a separate document.
10.
Action Item No.
10 of Bulletin 79-06A required that licensees review and modify, as necessary, maintenance and test procedures for safety-related systems 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.
In response to Bulletin Action Item No.
10, the licensee stated that the Ginna Technical Specifications require that pumps and valves of safety-related systems be tested monthly to check the operation of the starting circuits and to verify that the pumps are in satisfactory running order.
Rationale for selection of a one month test interval was provided.
If a component is found to be inoperable, every effort is made to effect repai rs and restore the system to full operability within a relatively short time.
To provide maximum assurance that the redundant component(s) will operate if required to do so, the redundant component(s) are tested prior to initiating repair of the inoperable component.
This applies to the safety injection pumps, residual heat removal pumps and their associated
- valves, and to the containment spray pumps and valves.
The two containment spray pumps must also be tested to demonstrate opera-bility before initiating maintenance on an inoperable charcoal filter unit.
With one diesel generator out of service, the remaining diesel generator is run continuously provided such operation does not exceed seven days (total for both diesels) during any one month, and provided that the station transformer is in service.
The Ginna Technical Specifications do not require testing of redundant components of the auxiliary feedwater system.
However, the, licensee revised the t1aintenance and Turbine Plant Operations Procedures to require testing of the redundant component(s) by the Results and Test Department prior to initiating repair of the inoperable component.
The procedures which were revised to include this requirement were identified.
13
n The licensee reviewed the Ginna maintenance procedures and determined that they were sUfficient to ensure the operability of safety-related systems after they are returned to service following maintenance.
After maintenance, the Results and Test Department performs the applicable periodic tests to assure that the system is operable.
Periodic test procedures were changed to require an additional verification by non-test personnel after testing of a safety-related system.
All safety-related valves manipulated during the test are checked to ensure that they are in their proper position.
After this verification, the completed test procedure must be reviewed by the Head Control Operator and approved by the Shift Foreman.
Ginna administrative procedures require that the Shift Foreman approve the removal from and the return to service of all safety-related systems.
On April 30,
- 1979, a procedural change was approved which required that the Mead Control Operator also be notified of such actions.
The transfer of information about the status of safety-related systems at shift change will be accomplished through Ginna Administrative Proce-dure A-52-1, "Shift Organization Relief and Turnov'er," which lists the l
requirements for shift turnover.
II Based on our review, we find the licensee's response to Bulletin Action I
Item No.
10 acceptable.
I'l.
Bulletin Action Item No.
I reporting procedures for notified within one hour l
or expected condition of continuous communication the NRC.
ll requested licensees to review their prompt NRC notification to assure that the NRC is of the time the reactor i's not in a controlled operation.
Further, at that time, an open, channel shall be established and maintained with In response to this bulletin action item,,the licensee implemented changes to the written plant procedures to notify NRC within one hour in the event that the reactor is not in a controlled or expected condition 14
of operation, such as that which would cause a valid safety injection actuation signal, or the necessity for site evacuation.
A telephone line located in the control room which was not used during normal operation was designated for open continuous communications after the initial notification to the NRC has been made.
Subsequent to the licensee's response to this bulletin action item, a
dedicated telephone line was installed for the purpose of establishing and maintaining the required communications channel.
The actions specified in Action Item No.
11 of IE Bulletin 79-OGA have the requirements of Section 50.72 of 10 CFR Part 50 immediately effective upon issuance February 29,1980.
Me find the licensee's response to Action Item No.
11 acceptable.
12.
In Action Item No. 12, licensees were requested to review operating
- modes, and procedures to deal with significant amounts of hydrogen gas that may be generated during a transient or other accident that would either remain inside the primary system, or be released to the containment.
In response to this bulletin action item, the licensee made a
comprehensive and knowledgeable evaluation of how hydrogen gas may be dealt with, both in the containment and in the primary coolant system.
The various methods which could be used for dealing with hydrogen in the primary coolant system were described.
A procedure was prepared and implemented in July 1979.
In addition to the above action, the licensee also participated, as a
member of the Mestinghouse Owners Group, in the effort to develop generic guidelines for emergency operating procedures in response to Item 2. l. 9 of NUREG-0578.
Included in this effort was the consideration of the treatment of non-condensible gas in the primary coolant system.
Our evaluation of Item 2. 1.9 implementation will be reported in a separate document.'
0, The licensee also reviewed the existing procedures for the removal of N
hyd'rogen from the containment, namely S.21.1 and S.21.2 (lA and 1B Hydrogen Recombiner, Purging and Operation, respectively).
These procedures'escribe the steps required to operate the hydrogen recom-biners to mai ntain the hydrogen concentration in the containment at a
safe level.
Based on our review, we find that the license has provided an adequate response to Bul'letin Action Item No.
12.
13.
This bulletin action item requested licensees to propose
- changes, as
- required, to those plant Technical Specifications which had to be modified as 'a result of'mplementing Bulletin Action Item Nos.
1 through 12, and to identify design changes necessary in order to effect long-term resolution of these items.
The licensee's response to this bulletin action item consisted of the June 7,
1979 submittal; which proposed changes to the Ginna Technic'al Specifications to accommodate the modification to the safety injection actuation logic previous'ly described in our evaldation of the licensee's response to.Bulletin Action Item No.
3.
lq We find the licensee's.
response to Bulletin Action Item No.
13 acceptable.
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, tl Revision 1.
The actions. taken demonstrate the licensee'-"s understanding of the concerns arising from the Three Nile Island, Unit No.
2 ~accident in relation to their implications of its own operations, provide added assurance for the protection of the public health and safety during plant 'operation.
~'I T.
This conclusion notwithstanding, it should be recognized that further actions will result from the staff's review of operating plants using nuclear steam 16
supply systems designed by Westinghouse (documented in NUREG-0611).
Additional changes have resulted from the staff's implementation of the requirements contained in NUREG-0578 (e.g.,
the actions taken in. response to Action Item 6 of Bulletin 79-06A regarding the PORVs).
Our evaluation of such matters will be provided in other documents.
17
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