ML18081B290
| ML18081B290 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 02/29/1980 |
| From: | Ross W Office of Nuclear Reactor Regulation |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8004020433 | |
| Download: ML18081B290 (14) | |
Text
Docket No. 50-272 MEMORANDUM. FOR.:
FROM:
SUBJECT:
fEBRUARY 2 9 1980 A. Schwencer, Chief, ORB #1, D9R William J. Ross? Project Manager, ORB #1, DOR
SUMMARY
OF DECEMBER 10, 1979 AUDIT OF IMPLEMENTATION OF SMALL BREAK lOCA GUIDELINES INTO PLANT PROCEDURES On December 10, 1979, members of the NRC staff met with representatives of Public Service Electric and Gas Company (PSE&G)*at the ~alem Generating Station site. The meeting was h~ld at the staff's request to al~ow the Bulleti~s and Orders Task Force to audit the implementation of the staff-approved Westinghouse generic guidelines for emergency procedures regarding small break lo~~~of-coolant accidents (LOCAs).
Attachment No. 2 is a list of participants. Attachment No. 1.
contains the agenda for the audit.
Before the audit started, the staff conducted a brief meeting with PSE&G repre-sentatives; P. 0' Reilly, B&O Task Force Project Manager for Westinghouse pl ants, (1) explained the purpose of the audit, (2) introduced the staff members present and the role that each would have during the audit, and (3) reviewed ~he audit agenda.
PSE&G then asked a number of questions about the scope of the audit.
One particular concern expressed by PSE&G was \\oJhether the audit would consider that impl~mentation of the generic guidelines and the appropriate.operator re-training were still in prqgress and had nqt yet been completed.
The staff assured PSE&G that it had considered the fact that the deadline for*implementation of the small break LOCA procedures, as specified in NUREG-0578., is January 1. 1980.
Consequently; the audit preparation had anticipated that the implementation of revised procedures and the corresponding operator retraining might not be completed.
The staff *assured PSE&G that, since tne audit was taking place before the imple-mentation date, it was not an 11 1nspfittion, 11 but an "impl en:ientation/ status/check.
11 PSE&G was enc~uraged to. fur~ish ~JlJf fee~back regarding )JnHJement~~ion, }P~,h: as problems putting the gu1del111es;>rnto the plant procedµres-; that m1g~e-useful on either a generic or a pla9t::.specific basis.
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PSE&G informed the staff that a problem had been encountered when the PSE&G operator trai~ing staff recently attempted to run the revised small break LOCA procedures on the Hesti.nghouse simulator at Zion, Illinois. It was discovered that, before all of the criteria required for termination* of high pressure safety injection (~~I) had been achieved, the pressurizer had filled with water r
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_..,..,_,..~' regarding the steam generator water level.
In the case of ~Jestinghouse-designed plants which were similar to the 412 standard plant (bigh-heat safety injection pumps:
~2500 psig shutoff head, the PORV would be challenged before the water level in the steam generator had reached the specified level.
PSE&G informed the staff that the Owners Group was working with Westinghouse to develop a solution to this problem.
The audit began with G. Kelly, M. Rubin, and B. Wilson 9f the B&O Task Force conducting a_ l ine-by-1 ine review with the PSE&G-representatives of the follow:-
ing Salem Unit No. 1 emergency proceaures:
I-4.0, 11Safety Injection Initiation,"
I-4.2, "Recovery from Safety Injection, 11 and I-4.4, "Loss of Coolant (Lea 1l~age
- Greater than Maximum Charging Fl ow),
11 against the* approved Westinghouse gu-1 de-1 ines* E-0, "Immediate Actions _anc! Diagnostics 11 and E-1, 11 Loss of Reactor Coolant.
11 Significant comments were noted for later discussion with PSE&G.
The next step in the audit consisted of a 11walk-through 11 of the above identified Salem Unit No. 1 emergency. procedures:
- l.
The "walk-through" was performed in the Salem Unit No. 2 control room.
The Unit 2 control room was used because it is almost identical to the Unit 1 control room and because its use would i~pact the work of the plant operations personnel less than if the Unit l control room were used. _ (Salem Unit No. 2 has not received an operating liGense).
- 2.
B.
~Jilson, B&O Task Force-, conducted this 11walk-through 11 with J. Bailey.
PSE&G. *At the same time*, G. Kelly and M.
Rubin~ B&O Task Force discussed various aspects of the procedure with a member of the PSE&G operations staff.
- 3.
Following *the procedural "walk-through, 11 B. Hilson, B~,Q Task Force, in- -
terviewed several -plant operators.
These interviews were conducted to deter-mine what operators had learned frdm their training since the Three Mile Island Unit 2 accident and what knowledge they had regarding the new small brea§ LOCA procedures.
- 4.
The NRC Resident Inspector {L. Norrholm) too~ a grdup of staff members
{A. Dromerick, G. Hollohan, P. O'Reilly, and W. Ross) on a "walk-trhough 11 of Units l and 2.
- 5.
G. Ke11y and M. Rubin, B&O Task Force discussed the design of the contain-ment sump and the-staff's requirements regarding ECCS performance testing with PSE&G.
Following the audit, a brief meeting was held between the staff and PSE&G to summarize the staff's findings during the audit.
B. Wilson, B&O Task Force, informed PSE&G that the interviews with theplant operators revealed the following:
(1) the operators evfdenced some confusion about the exact sequence
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He discussed operating.license examination procedures.
He suggested that the training program place increased emphasis on certain aspects of thermo-dynamics, such as1 two-phase and saturated conditions, heat transfer, and fluid flow.
PSE&G commented that the training program personnel would soon be re-porting to offsite management instead of to the Plant Superintendant.
Eventually, training offices* will be located at a new offiflte PSE&G Simulator Center.
The staff expressed c6ncern with this proposal.
G. Kelly and M~ Rubin, B&O-Task Force, presented a number of comments which they recommended that PSE&G act on, either by revising the small break LOCA procedures, or by other appropriate implementation of the prescribed action.
These are identified in Enclosure No. 3.
The method to be used for followup on the staff's comments was then discussed.
Since some uncertainty existed about this matter, the staff informed PSE&G that further guidance would be forthcoming on how this followup i-muld be done.
(Subsequent to the site visit, the staff made the decision that the followup review will be performed by the Office of Inspection and Enforcement through the Resident Inspector where possible.
Eoclosures:
- l. Audit of Small Break LOCA Procedures
- 2.
List of Attendees
- 3.
Specific Comments
/
/'
/
- f_
- /
William J. Ross, Project Manager Op~rating Reactors Branch No. 1 Division of Operating Reacto~s
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Meeting Summary for Sal em l Docket Fil es NRC PDR Local PDR OR~l Reading NRR Reading
- H. Denton E. Case D. Ei senhut.
R. Tedesco G. Zech B. Grimes W. Gammill L. Shao J. Mi 11 er R. Vollmer T. J
- Ca rte r
- A. Schwencer D. Ziemann P. theck
- G. Lainas D. Crutchfi e 1 d i3. Grimes T. Ippolito R. Reid
- v. Noonan
- G. Knighton
- o. Brinkman Project Manager OELU OI&E (3)
- c. Parrish/P. Kreutzer ACRS. (16)
NkC Pa rt i ci pants NSIC TERA Licensee Short Serv.i ce Li st 4; -
Mark J. Wetterhahn, Esquire Conner, Moore and Corber Suite l 050 1747 Pennsylvania Avenue, NW Washington, D. C.
20006 Richard Fryling, Jr., Esquire Assistant General Solicitor Public Service Electric and Gas 80 Park Place Company 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. Mitt1, General Manager Licensing and Environment Public Service Electric and Gas Company 80 Park Place Newark, New Jersey 07101 Salem Free Library 112 West Broadway Sal em, New Jersey 08079 Leif J ~ Norrholm U. S. Nuclear Regulatory Commission Drawer I*
Hancocks Bridge, New Jersey 08038
- -1
I.
Docket No. 50-272 MEMORANDUM FOR:
FROM:
SUBJECT:
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 fESRUARY 2 9 1980 A. Schwencer, Chief, ORB #1, DOR William J. Ross, Project Manager, ORB #1, DOR
SUMMARY
OF DECEMBER 10, 1979 AUDIT OF IMPLEMENTATION OF SMALL BREAK LOCA GUIDELINES INTO PLANT PROCEDURES On December 10, 1979, inembers of th.e NRC staff met with representatives of
- Publi~* Service Electric and Gas CompanY (PSEgG) at the Salem Generating ~tation site. The meeting was held at the staff's request to allow the Bulletins and Orders-Task Force to audit the implementation of the staff-approved Westinghouse generic guidelines for emergency procedures regarding small break loss-of-coolant acciden~s (LOCAs).
Attachment No. 2 is a list of participants. Attachment No. l contains the agenda for the audit.
Before the audit started, the.staff conducted a brief meeting with PSE&G repre-sentatives.
P. O'Reilly, B&O Task Force Project Manager-for Westinghouse plants, (1) explained the purpose of the audit, (2) introduced the staff members present and the *role that each would have during the audit, and (3) reviewed the audit agenda.
PSE&G then asked a number of questions about the scope of the audit.
Dne particular concern expressed by PSE&G was whether the audit would consider that implementation of the generic guidelines and the appropriate operator re-training were still in progress and had not yet been completed.
The staff assured PSE&G that it had considered the-fact that the deadline for implementation of the small break LOCA procedures, as specified in NUREG-0578, is January.l, 1980.
Consequently, the audit preparation had anticipated that the implementation of revised procedures and the corresponding operator retraining might not be completed.
The staff assured PSE&G that, since the ~udit was taking place before the imple-mentation date, it was not an "inspection," but an "implementation status check."
PSE&G was encourqged to furnish any feedback regarding implementation, such as pro~lems putting the guidelines into the plant procedures, that might be useful on either a generic or a plant-specific basis.
PSE&G informed the staff that a problem had been encountered when the PSE&G operator training staff recefitly attempted to run the revised small break LOCA procedures on the Westinghouse simulator at Zion, Illinois. It was discovered that, before all of the criteria reciuired for termination of high pressure safety injection (HPI) had been achieved, the pressurizer had filled with water and the pressurizer power-operated relief valve (PORV) had been actuated to provide pressure relief.
Discussions between PSE&G and the Westinghouse Owners Group led to the discovery that at-least one other licensee had expe.rienced the same problem.
The Owners Group had discussed this matter with Westinghouse.
The cause of the problem had been identified as the HPI termination criterion
.. regarding the steam generator water level.
In the case of Westinghouse-designed plants which were similar to the 412 standard plant (high-heat safety injection pumps:
~2500 psig shutoff head} the PORV would be challenged before the water level in the steam generator had reached the specified level.
PSE&G informed the staff that the Owners Group was working with Westinghouse to develop a solution to this problem.
The audit began with G. Kelly, M. Rubin, and B. Wilson of the B&O Task Force conducting a line-by-line review with the PSE&G representatives of the follow--
ing Salem Unit No. l emergency procedures:
I... 4.0, 11Safety Injection Initiation, 11 I-4.2, 11Recovery from Safety Injection, 11 and I-4.4, 11 Loss of Coolant (Leakage Greater than Maximum Charging Flow),
11 against the approved Westinghouse guide-1 ines E-0, 11 Immediate Actions and Diagnostics 11 and E-1, 11 Loss of Reactor Coolant.
11 Significant comments were rioted for later discussion with PSE&G.
The next step tn the audit cpnsisted of a 11walk..,through 11 *of the above identified Sil em Unit No.* 1 emergency procedures:
- 1.
The 11walk-through 11 was perfo.rmed in the Salem Unit No. 2 control room..
The Unit 2 control room was used because it is almost identical to the Unit 1
- control room and because its use would impact the work of the plant operations personnel less than if the Unit l control room were used.
(Salem Unit No. 2 has not received an operating license).
- 2.
B. Wilson, B&O.Task Force, conducted this 11walk-through 11 with J. Bailey.
PSE&G.
At the same time, G. Kelly and M. Rubin, B&O Task Force discussed various aspects of the procedure with a member bf the PSE&G operations staff.
-3.
Following the procedural 11walk-through, 11 B. Wilson, B&O Task Force, in-terviewed several plant operators. These interviews were conducted to deter-mine what operators had learned from their training since the Three Mile Island Unit 2 accident and what knowledge they had regarding the new small break LOCA procedures.
- 4.
The. NRC Resident Inspector (L. Norrholm) took a group of staff members (A. Dromerick, G. Hallahan, P. O'Reilly, and W. Ross) on a 11walk--through 11 of Units 1 and 2.
- 5.
G. Kelly and M. Rubin, B&O Task Force discussed the design of the contain-ment sump and the staff's requirements regarding ECCS performance testing with PSE&G.
Following the audit, a brief meeting was held between the staff and PSE&G to summarize the staff's findings during the audit.
B. Wilson, B&O Task Force, informed PSE&G that the interviews with the plant operators revealed the following:
(1) the operators evidenced some confusion about th_e exact sequence of events during the early part of the TMI-1 accident, and (2) some operators needed more background on the behavior of reactors under saturated conditions.
He discussed operating license examination procedures.
He suggested that the training program place increased emphasis on certain aspects of thermo-dynamics, such as two-phase and saturated conditions, heat transfer, and fluid flow.
PSE&G commented that the training program personnel would soon be re-porting to offsite management instead of to the Plant Superintendant.
Eventually, training offices will be located at a new offsite PSE&G Simulator Center.
The staff expressed concern with this proposal.
G. Kelly and M. Rubin, B&O Task Force, presented a number of comments which
. they recommended that PSE&G act on, either by revising the small break LOCA procedures, or by other appropriate implementation of the pre.scribed action.
These are identified in Enclosure No. 3.
The method to be used for followup on the staff's comments was then discussed.
Since some uncertainty existed about this matter, the staff informed PSE&G that further guidance would be forthcoming on how this followup would be done.*
(Subsequent to the site. visit, the si;aff made the-decision that the followup review wili be performed by the Office* of.Inspection and Enforcement through the Resident Inspector where possibleJ
Enclosures:
- 1. Audit of Small Break LOCA Procedures
- 2.
List of Attendees
- 3.
SpecifJc Comments William J. Ross, Project Manager Operating Reactors Branch No. 1 Division of Operating Reacto~s
Attachnent 1 AUDIT OF SMALL BREAK LOCA PROCEDURES
- 1.
Comparison to Guidelines The NRC staff will compare the small break procedure to the guidelines approved for the.plant type.
In addition to checking against the specific line itetIS (actions and precautions) in the guideline, the NRC staff will consider the clarity of the procedure, in terms of individual actions and precautions, and flow of the procedure with respect to timely initiation of all operator actions.
The licensee should have personnel familiar* with *the *development of the sirall break LOCA procedure available to answer NRC staff questions about omissions and.ambiguities in*the procedure.
The utility should explain the process by which the approved guidelines were converted into operating procedures including the' results of internal review, problem areas. and plant specific information required in the procedure.
- 2.
Training The NRC staff will review the training the reactor ~perators received in the small break* LOCA procedure.
We expect this training to include fonnal classroom study and a walk thru of the procedure with the shift supervisor or training *coordinator.
The licensee should have the training coordinator available to discuss the training with the NRC staff.
- 3.
Operator Feedback A member of the Operator Licensing 3ranch or an IE Inspector will audit several operators to-determinP the acecuacy of the procedure from a
i func:t.ional sr:andpoirit and tht! ~ffectiveness of the training program..
- The audit,. which is oral, will cover understanding of small break LOCA, differ-
- erntiation between LOCAs and other depressurization events, familiarity with the sma 11 break LOCA procedures, hasi s for the revised procedures,._ and whether the procedures can be effectively carried out.
The licensee should have several operators available, including senior*
operators* who can be questioned by the staff.
- 4. Svstems Considerations
- rne NRC staff will review system-related aspects of the procedure to assure that the operator actions can be perfonne-0. These Systems considerations wil 1 vary rli 'th p 1 ant type as noted in Tab 1 e 1.
The l i censt:e shoLil d have techni ca 1 personne1 and docurnentati on avai 1 able
_(including P&ID of ECCS) to respond to NRC staff questions on system considerations in the procedure.
.. __.... \\~es ti nghouse a)
Instrumentation Certain instrumentation is used by the operator in the procedure to initiate actions such as HPI tennination and switchover from injection to recirculation. The licensee. should show that the instrumentation used as part of ttiis procedure wi11 provide adequate information for perfonning operator.actions considering instrument uncertainty, environmental conditions at the time it is required, and power (with loss of offsite power and a single failure).
b)
Reactor Coolant-Pump Trip - The licensee s*hould show that the primary system pressure at which he trips the reactor coolant pumps in his procedure is consistent with the prescription in the guidelines.
c)
Manual Switchover - For those plants which rely on the operator to perform a portion or all of the switchover from injection to recirculation,
- the 1 i censee should show that this acti ori can be performed before the RWST water level reaches an unacceptable level. A time sequence table should be available showing procedure steps, flow out of the RWST, volume of water remaining in the RWST, and time required.to perform each step as a function of time.
Maximum ECCS and containment spray flmv should be used to develop the sequence table.
d)
HPI Protection After Switchover - For those plants that have HPI pumps with 1400-1500 psi shut off heads, the licensee should show how the procedures will preclude damage to those pumps after switchover to recirculation from the sump for break sizes that would result in deadheading the pumps after switchover.
Under these conditions, the procedures shoulq also provide instructions for maintaining adequate pressure level inventory
'. after switchover.
- e. Bulletin Items - The licensee shculd show how he has incorporated his responses to the I&E bulletins (issued in April 1979) in his procedure with respect to:
Containment Isolation PORV Indication Transfer of Fluids Out of Contairn:ent Hydrogen Gas ATTENDEES NRC PSE&G P. D. 0' Reilly R. W. Skwarek
- w. J. Ross G. w. Kapp B. A. Wilson J. v. Bailey G. M. Holahan J. M. Zupko A. W. Dromerick
. M. P. Rubin.
G. B. Kelly L. J. Norrholm SPECIFIC COMMENTS
- 1. Operator licensing should emphasize the following subjects:
a) thermodynamics b) heat transfer c) fluid flow
- 2.
There was apparent confusion about the behavior of water under two-phase conditions.
- 3.
Operators had the misconception that, in the TMI-2 sequence of events, saturation was reached immediately thereby causing loss of pressurizer level while, actually, saturation was not reached until seven minutes into the accident after loss of the heat sink and partial termination of HPI.
- 4.
Future training will _be *performed at a Simulator Center that will be located in another part of New Jersey.
- 5.
fhe procedures need a note related to manually loadi~g buses, i.e.,
sequence of loading and time between attempts to load equipment on buses.
- 6.
PORVs normal air supply is isolated on Phase A.
Containment isolation can go 80 cycles on bottled air.. (I-4.0)
- 7.
Steps 3.2 and 5.3 in the Emergency Procedure conflict on reactor coolant pump trip pressure.
- 8.
The immediate actions do not have checks for:
a) bus voltages and loading b) auxiliary feedwater* flow c) component cooling water flow I-4.0 I-4.0 I-4.0
. 9.
The immediate actions do* not contain discussions on Containment Hi-2 and Hi-3 (see C#4 and #5 in Westinghouse guidelines)
I-4.0J
- 10.
Step 5.1 (I-4.0) should not be limited to cases where there is no other leak indications.
- 11.
How is the PORV determined to be closed? -
- 12.
Caution is required to tri~ the Reactor Cooling Pumps (RCPs) if component cooling water is lost for five minutes.
If RCPs are tripped, cooling water must continue to be supplied to the pump seals.
- 13.
I.4-2 step 4.1 needs caution on manual reload of loads on diesels and reference for load sequence.
- 14.
There is no discussion in I-4.4 of inadequate core cooling.
9Attachment 3 15. After trip of RCPs there is no verification of notice (I-4.0, I-4.4)
- 16.
Need cautions in I-4.0 for RWST level.
- 17.
There is no discussion of pressurizer level in I-4.0 and I-4.4.
- 18.
No safety injection pump miniflow protection after miniflow valves are isolated..
- 19. There is only a single wide-range-level indication per steam generator.
These*indicators are not on the main control board and not easily visible.
All four indicators are on one nonvital bus panel.
- 20.
11Sufficient NPSH 11 light in step 5.2 of I-4.4 is not satisfactory. There should either be instructions on what to do if this light is not on (or if a light is glowing on the next lower light level) or the operator should depend only on the RWST level. The NPSH light and switch are
-not environmentally qua.lified for a LOCA.
. 21.
The direct readout range of the 11Temperature Readout 11 -wheel must extend from 700°F to at least l000°F.
- 22.
The hot leg resistance temperature devices (RTDs) are on non-vital buses.