ML19260C055
| ML19260C055 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 11/23/1979 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| IEB-79-05A, IEB-79-05B, IEB-79-5A, IEB-79-5B, NUDOCS 7912180287 | |
| Download: ML19260C055 (17) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION h
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t!ovember 23, 1979 dj Docket No. 50-312 Mr. J. J. Mattimoe Assistant General Panager and Chief Engineer Sacramento Municipal Utility District 6201 S Street P. O. Box 15830 Sacramento, California 95813
Dear Mr. Mattimoe:
We have reviewed the information provided by your letters dated April 11, 16, 19 and 22 and May 2, 11, 14 and 21 and October 26, 1979 in response to IE Bulletins79-05A and 75-058 for the Rancho Seco t!uclear Generating Station.
The enclosure provides an evaluation of your responses and dis-cusses them with respect to their specificity, completeness and responsive-ness to the intent of the requirements of these bulletins.
We have found that you have taken the appropriate actions to meet the majority of the requirements of IE Bulletins79-05A and 79-053. With regard to the outstanding items identified in our evaluation, these will be dealt with in the continuing staff review of the Three Mile Island, Unit No. 2 accident and other corrective actions may be required at a later date.
In this regard, IE Bulletin 79-05C was issued on July 26, 1979 requiring new considerations for operation of th? reactor coolant pumps following an accident. Our review of your responses to IE Bulletin 79-05C dcted August 27, September 19 and October 24, 1979 is continuing pending your response to short-term action item 5 and long-ter n action item 1 of tne bulletin.
Sincerely, G..L:. / u.1M m
Robert W. Reid, Chief Operating Reactors Branch #4 Division of Opera-ing Reactors
Enclosure:
Evaluation of Licensee's Res::onses to IE Bulletins79-05A and 79-052
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See next : age m
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Sacram nte Municipal Utility District cc w/ enclosure (s):
Christopher Ellison, Esq.
David S. Kaplan, Secretary and Dian Grueuich, Esc.
General Counsel California Energy Commission 6201 5 Street 1111 Howe Avenue P. O. Box 15830 Sacramento, California 95825 Sacramento, California 53813 Ms. Eleanor Schwart::
Sacramento County California State Office Board of Supervisors 600 Pennsylvania Avenue, S.E., Rm. 201 827 7th Street, Room 424 Washington, D.C.
20003 Sacramento, California 95814 Docketing and Service Section Office of the Secretary usiness and Municipal Department U. S. Nuclear Regula ory Commission m
Sacramento City-County Library Washington, D.C.
20555
.28 I Street Sacramento, California 95814 Michael L. Glaser, Esq.
1150 17th Street, N.W.
Director, Technical Assessment Washington, D.C.
20036 Divisien Office of Radiation Programs Dr. Richard F. Cole (AW-459)
Atomic Safety and Lic' nsing Board U. S. Environmental Protection Agency Panel Crystal Mall #2 U. S. Nuclear Regulatory Commission Arlington, Virginia 20460 Washington, D.C.
205c:
U. S. Environmental Protection Agency Mr. Frederick J. Shon Region IX Office Atomic Safety and Licensing Board ATTN:
EIS COORDINATOR Panel 215 Fremont Street U. S. Nuclear Regulatory Commission San Francisco, California 94111 Washington, D.C.
20555 Mr. Robert B. Borsum Timothy V. A. Dillon, Esq.
Sabcock & Wilcox Suite 380
'Scisar Power Generation Division 1850 K Street, N.W.
evite J20, 7735 Old Georgetown Road Washington, D.C.
20006 Lotnesda, Maryland 20014 James S. Reed, Es Thomas Baxter, Esq.
Michael H. Remy, g.
Shaw, Pittman, Potts & Trowbridge
- sq.
1800 M St Reed, s,amuel & Remy Washington,'iW DC 20036 717 K Street, Suite 405 Sacramento, California 9581a Heroert H. Brcwn, Esq.
Mr. Micnael R. Eaton Lawrence Coe Lanpher, Esq.
Energy Issues Coordinator Hill, Christopner and Phillips, P. C.
Sierra Club Legislative Office 1900 M St., NW 1107 9th St., Room 1020 Wasnington, D. C.
20036 Sat.amento, CA 95812 B
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Sacramento Municipal Utility District cc w/ enclosure (s):
Atomic Safety and Licensing Board Panel V. S. Nuclear Regulatory Commission Washington, D.C.
20555 Atomic Safety and Licensing Appeal Board Panel U. S. Nuclear Regulatory Commission Washington, D.C.
20555 fir. Richard D. Castro 2231 K Street Sacramento, California 95814 Mr. Gary Hursh, Esq.
520 Capital Mall Suite 700 Sacramento, California 95814 California Department of Health ATTN: Chief, Environmental Radiation Control Unit Radiological Health Section 714 P Street, Room 498 Sacramento, California 95814
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Inclosure EVA'.UATION OF LICENSEE'S RESPONSES TO IE BULLETINS79-05A AND 79-058 SACRAMENTO MUrICIPAL UTILITv DISTRICT RANCHO SECO NUCLEAR GENERATING STATION DOCKET NO. 50-312 b
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. INTRODUCTION On March 28, 1979 the Three Mile Island Nuclear Power Plant, Unit 2 (THI-2) experienced core damage which resulted from a series of events which were ini-tiated by a loss of feedwater transient.
Several aspects of the accident have generic applicability at operating Babcock and Wilcox (B&W) reactors.
On April 1, 1979, IE Bulletin 79-05 was sent to all B&W operating plant licensees.
The purpose of the bulletin was to provide information concerning the accident at THI-2 and to request certain actions be taken by licensees to preclude a similar occurrence at their facilities.
This bulletin was superseded and expanded by IE Bulletin 79-05A dated April 5,1979, and by IE Bulletin 79-05B dated April 21, 1979.
By letters dated April 11, 16, 19 and 22 and May 2, 11, 14 and 21 and October 26, 1979, the Sacramento Municipal Utility District (SMUD or licensee) provided responses in conformance with the requirements of the bulletins.
Information became available to the NRC, subsequent to the issuance of IE Bulletin 79-05B, which required modification to item 4.c of IE Bulletin 79-05A.
On July 26, 1979, IE Bulletin 79-05C'was issued superseding item 4.c of IE Sulletin 79-05A. A separate evaluation of responses to IE Bulletin 79-05C'is presently being conducted and will be published at a later date.
Subsequent to the issuance of IE Bulletins 79-05,79-05A, and 79-05B., the Commission issued an Order dated May 7, 1979, which confirmed the licensee's commitment to make certain modifications to plant equipment and procedures, and to complete specific operator training and analyses of plant behavior.
Due to the overlap in the requirements of the Order and the bulletins, the Order is referenced several times in this evaluation.
In addition, the NRC's Lessons Learned Task Force has recently completed its report (NUREG-0578) detailing short-term recommendations that are to be implemented for all
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16M 197
. operating reactor plants in light of the accident at THI-2. This report has also been referer.ced in this evaluation.
The NRC staff's evaluation of the licensee's compliance with the short-term portion of the Commission Order was issued on June 27,1979 (Reference 121 Separate evaluations of the licensee's compliance with the long-term portion of the Comission Order and NUREG-0573 will be issued at a future date.
The NRC staff's evaluation of the licensee's responses to IE Bulletins79-05A and 79-05B is provided below.
Certain items in this evaluation will require further staff review during its evaluation of the licensee's compliance with the long-term portion of the Commission Order and the licensee's implementation of NUREG-0578. Where applicable, these issues are discussed under the appro-priate bulletin item and a summary of the outstanding issues is provided at the end of this evaluation.
EVALUATION OF RESPONSES TO IE BULLETIN 79-05A
, Item 1 "In addition to the review of circumstances described in Enclosure 1 of IE Bulletin 79-05, review the enclosed preliminary chronology of the THI-2, 3/28/79, accident.
This review should be directed toward understanding the sequence of events to ensure against such an accident at your facility (ies)."
The licensee has reviewe_ fnclosure I to IE Bulletin 79-05 and the preliminary sequence of events enclosed with IE Bulletin 79-05A.
The licensee, assisted by B&W, assessed the adequacy of Rancho Seco to safely sustain transients such as the one which occurred at TNI-2.
Its rr,iew identified the same six human, design and mechanical failures which resulted in the core damage and radiation releases at TMI-2, as are described in the " Description of Circumstances" portion of IE Sulletin 79-05A.
Details of this review are documented in Reference 1 to this evaluation.
The staff has reviewed this document and
. finds the licensee has a satisfactory understanding of the sequence of events.
Specific staff comments concerning the licensee's response are provided below.
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. The description of circumstances in IE Bulletin 79-05A states that the pres-surizer electromatic relief valve (PORV), which opened during the initial pressure surge, failed to close when the pressure decreased below the actuation level.
In Reference 1, the licensee stated that at Rancho Seco, a temperature monitor downstream of the PORV is alarmed to provide indication in the control room that this valve is open. We believe that the temperature monitor alone is not always a valid indication of PORV position, and that a more direct means of monitoring the PORV position should be available. This item will be resolved as part of the implementation of Section 2.1.3a of NUREG-0578.
In the interim, the staff has reviewed the licensee's emergency procedure for
" Loss of Reactor Coolant / Loss of Reactor Coolant Pressure" and verified that the leak isolation section of the procedure requires the operator to isolate the PORV by shutting the block valve upstream of the PORV.
The description of circumstances in IE Bulletin 79-05A states that because the containment did not isolate on high pressure injection (HPI) initiation, the highly radioactive water from the PORV discharge was pumped out of the contain-ment by the automatic initiation of a transfer pump.
The licensee stated in Reference 1 that at Rancho Seco the containment isolation system is initiated by a 4 psig high reactor containment pressure signal or by a 1,600 psig low reactor primary coolant system pressure.
The licensee further stated that the safety features initiation of reactor containment isolation would isolate the Rancho Seco containment early in a transient, well before any significant release of radioactive material would occur.
Item 9 of this bulletin requires that the licensee review operating modes and procedures to further assure that undesired pumping of radioactive liquids and gases out of the containment will not occur inadvertently. The issue of containment isolation is addressed in our evaluation of the response to item 9.
The description of circumstances in IE Bulletin 79-05A discusses the adverse effect of intermittent operation of the HPI system at TMI-2.
The licensee stated in Reference 1 that recent revisions to the operating and emergency procedures at Rancho Seco, as required oy items 4a, b and d of IE Sulletin 79-05A, will preclude occurrence of a similar event at Rancho Seco.
This matter is accressed in our evaluation of the responses to t
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The description of circumstances in IE Bulletin 79-05A states that tripping of the reactor coolant pumps (RCPs) during the course of the transient, to protect against RCP damage due to vibration, led to fuel damage since the core was uncovered and voids in the reactor coolant system (RCS) prevented natural circulation.
A-discussed in References 1 and 2, the licensce had modified its procedures to assure at least one RCP per loop remained operating during a loss of reactor coolant / loss of reactor coolant system pressure transient, in accordance with NRC guidance in item 4.c of IE Bulletin 79-05A.
As discussed in the introduction to this evaluation, this requirement has been modified and superseded by the requirements of IE Bulletin 79-05C.
This issue is also discussed under item 4.c of IE Bulletin 79-05A.
The NRC staff finds that the licensee has been responsive to item 1 of IE Bulletin 79-05A.and that any further follow-up action on diaect PORV position indication will be handled under Section 2.1.3a of NUREG-0578.
Therefore, the NRC staff considers the licensee's response to this item complete.
Item 2:
"Revi,ew any transients similar to the Davis-Besse event (Enclosure 2 of IE Bulletin 79-05) and any others which contain similar elements from the enclosed chronology (Enclosure 1) which have occurred at your facility (ies).
If any significant deviations from expected performance are identified in your review, provide details and an analysis of the safety significance together with a cescription of any corrective actions taken.
Reference may be made to previous information provided to the NRC, if appropriate, in responding to this item."
In response to item 2 of Bulletin 79-05A, the licensee stated in Reference 1 that it had reviewed transients which had occurred at Rancho Seco to determine if any had elements similar to the chronology of events at TMI-2 (March 28, 1979) and Davis-Besse 1 (November 29, 1977).
Based upon this review, the licensee stated tnat it had not icentified any transients which were similar; however, it had reviewed one transient in which the cooldown resulted in operation outside the Technical Specification pressure /temoerature limits.
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16N 200 This was a cooldown transient which was reported as Reportable Occurrence (RO) 78-01 dated March 31, 1978 (Reference 10). The initiating event for this transient was an inadvertent loss of power to the non-nuclear instrumentation (NNI) which provides signals to the integrated control system (ICS).
As a result of this event, the licensee developed an emergency procedure that allows plant conditions to be stabilized follo ing a loss of NNI power.
An additional transient, not identified by the licensee in its bulletin response, that resulted in less severe cooldown of the reactor coolant system than that identified in RO 78-01, was described in the licensee's letter dated January 25, 1979 (Reference 11).
In this case, the initiating event was a loss of the ICS power supply.
During the staff's on going evaluation of the TMI-2 accident, the transients noted above will be reviewed to determine whether further changes or modifica-tions may be desiracle to give added assurance that a TMI-2 accident will not be repeated.
In particular, the Commission's Order of May 7, 1979 required the licensee to submit a failure modes and effects analysis of the ICS.
This report was submitted on August 17, 1979 (Reference 15) and is presently under joint review by the NRC staff and the Oak Ridge National Laboratory.
The NRC staff considers the licensee's response to item 2 of IE Bulletin 79-05A complete.
Item 3:
" Review the actions required by your operating procedures for coping with transients and accidents, with particular attention to:
a.
recognition of the possibility of forming voids in the primary coolant system large enough to compromise the core cooling capability, especially natural circulation capability; b.
operator action required to prevent the formation of sucn voids; and 0
1607 201 c.
operator action required to enhance core cooling in the event such voids are formed."
As a result of the THI-2 accident, all licensed operators have received additional training to enable them to cope with transients and accidents. The specific training received to comply with subparagraphs a, b and c above are documented ir. Reference 1 to this evaluation.
Specific staff comments on this training are provided below.
To accomplish this training, the licensee stated all licensed operators would receive a training program using the B&W simulator, which would demonstrate the TMI-2 event and actively engage them in the analyzing and correcting abnormal transient situations including steam void formation in the reactor coolant system. The licensee stated that this training would be completed within a 120-day period. The staff considered the length of time for comple-tion of this training was unduly protracted.
However, simulator training for all licensed operators was required by tne Commission's Order of May 7,1979 and this training was completed for all licensed operators on June 21, 1979.
Reference 12 documents the NRC staff's evaluation of this training.
As part of the response to subparagraph c above, the licensee stated that it had modified operating procedures, as required in item 4.b of IE Bulletin 79-05A, to assure continued operation of at least one reactor coolant pump per loop in an emergency to assist in core cooling during accident conditions.
As discussed previously in this evaluation, IE Bulletin 79-05C requires licensees to immediately trip all operating reactor coolant pumps in the event of a reactor trip and initiation of HPI caused by reactor coolant system low pressure.
Rancho Seco operating procedures have been modified to reflect the requirements of IE Bulletin 79-05C.
Additional requirements in the area emergency procedures for transients and accidsnts have been recommended in Section 2.1.9 of NUREG-0578.
To ccmply with these requirements, the licensee is actively engaged in developing operator guidelines which cover inadequate core cooling and other abnormal transients.
O 16M 202 The schedules for completing these items are also found in NUREG-0578.
These requirements greatly expand the actions required by item 3 of this bulletin.
The NRC staff considers the licensee's response to item 3 of IE Bulletin 79-05A complete.
Item 4:
" Review the actions directed by"the operating procedures and training instructions to ensure that:
a.
operators do not override automatic actions of engineered safety features; b.
operating procedures currently, or are revised to, specify that if the high pressure injection (HPI) system has been automatic-ally actuated because of low pressure condition, it must remair, in operation until either; (1) both low pressure injection (LPI) pumps are in operation and flowing at a rate in excess of 1000 gpm each and the situation has been stable for 20 minutes, or (2) The HPI system has been in operation for 20 minutes, and all hot and cold leg temoeratures are at least 50 degrees below the saturation temperature for the existing RCS pressure.
If the 50 degree subcooling cannot be maintained after HPI cutoff, the HPI shall be reactiviated; c.
operating procedures currently, or are revised to, specify that in the event of HPI initiation, with reactor coolant pumps (RCP) operating, at least one RCP per loop shall remain operating.
d.
operators are provided additional information and instructions to not rely upon pressuri::er level indication alone, but to 6
16M 203 also examine pressurizer pressure and other plant parameter indications in evaluating plant conditions, e.g., water inventory in the reactor primary system."
IE Bulletin 79-058 modified the actions required in subparagraphs a and b above to take into account pressure vessel integrity considerations.
Evaluation of this matter is discussed under item 2 of IE Bulletin 79-05B.
IE Bulletin 79-05C modified the actions required in subparagraph c above.
The licensee's response to IE Bulletin 79-05C is presently undergoing staff review.
A separate document will be published in the near future which will present the staff's evaluation of all pressurized water reactor licensee's responses to IE Bulletins79-05C and 79-06C.
In regard to subparagraph d above, the licensee has documented in Reference 2 to this evaluation that all licensed operators have been given direction to utilize the pressure / temperature relationship of the reactor coolant system to assure proper subcooling prior to securing HPI. Guidance is provided to the operator in the control room in the form of a pressure versus temperature graph that clearly shows regions where the reactor coolant system is in a saturated condition and where it is 50 degrees subcooled. We have also reviewed the licensee's procedure for loss of coolant and consider that adequate guidance is given to the operator and that many indications, not just pressurizer level alone, are available to assist the operator in assessing the reactor coolant system water inventory.
In addition, Section 2.1.9 of NUREG-0578 requires that licensees upgrade reactor instrumentation to provide the operator with an unambiguous indication of vessel water level and core cooling adequacy.
The NRC staff finds that the licensee has been responsive to item 4 of IE Sulletin 79-05A and that any further follow-up action on item 4.c will be handlea under the NRC staff review of IE Bulletin 79-05C.
Therefore, the NRC staff considers the licensee's response to this item complete.
O 16M 204 Item 5:
" Verify that emergency feedwater valves are in the open position in accordance with item 8 below.
Also, review all safety-related valve positions and positioning reqtrirements to assure that valves are positioned (open or closed) in a manner to ensure the proper operation of engineered safety features. Also review related procedures, such as those for maintenance and testing, to ensure that such valves are returned to their correct positions following necessary manipulations."
The licensee has documented in Reference 1 to this evaluation, that procedures providing valve line-ups for engineered safety features have been reviewed and that valve positions have been verified against these procedures except where entry into the reactor containment was required. The valve position verifica-tion requiring reactor containment entry was accomplished during the shutdown following issuance of the Commissions's Order of May 7,1979.
A review of related maintenance and testing procedures was also completed by the licensee.
This matter is more fully discussed under item 10 of this Bulletin.
The NRC staff finds the licensee's response to item 5 of IE Bulletin 79-05A complete.
Item 6:
" Review the containment isolation initiation design and procedures, and prepare and implement all changes necessary to cause containment isolation of all lines whose isolation does not degrade core cooling capability upon automatic initiation of safety injection."
The licensee has documented in Reference 2 to this evaluation, that contain-ment isolation at Rancho Seco occurs at either 1,600 psig reactor coolant system pressure or at 4 psig reactor containment pressure. The licensee has stated that it has reviewed containment isolation design and procedures and has found them acceptable. All lines not required for safety features are isolated upon initiation of containment isolation except lines that are necessary to assure continued operation of the reactor coolant pumos and control rods.
The acceptability of not isolating the lines for reactor coolant pumps and control rod operation will be reviewed as part of the licensee's compliance 16 205 with Section 2.1.4 of NUREG-0578.
D M
k The NRC staff finds that the licensee has been responsive to item 6 of IE Bulletin 79-05A and that any further resolution of the items discussed above will be handled under Section 2.1.4 of NUREG-0578.
Therefore, the NRC staff considers the licensee's response to this item complete.
Item 7:
"For manual valves or manually-operated, motor-driven valves, which could defeat or compromise the flow of auxiliary feedwater to the steam generators, prepare and implement procedures which:
a.
require that such valves be locked in their correct position; or b.
require other similar positive position controls."
The licensee has documented in Reference 2 to this evaluation that surveillance procedures have been established and reviewed and that valve positions have been verified to be corre;t for the manual and manually-operated, motor-driven valves which could defeat or compromise the flow of auxiliary feedwater to the steam generators.
Each manually-operated valve in the system is locked open as required by Surveillance Procedure SP214.03 (" Locked Valve List"). The motor-operated valves are closed and stroked quarterly to verify operability.
These valves open on a safety features actuation signal.
Position status for all the motor-operated valves in the system is indicated in the control room.
Although not specifically addressed in the bulletin, the staff was concerned about the operation and reliability of the two pneumatically-operated flow control valves in the Rancho Seco auxiliary feedwater system.
These valves are controlled by the ICS.
As part of complying with the immediate actions of the Commission's Order of May 7,1979, the licensee developed a procedure that allows operator control of the auxiliary feedwater flow independent of the ICS through the safety grade bypass valves.
This action has resolved the staff's concern about the pneumatically-operated valves.
li 1607 206 The NRC staff finds the licensee's response to item 7 of IE Bulletin 79-05A complete.
Item 8:
" Prepare and implement immediately procedures which assure that two independent steam generator auxiliary feedwater flow paths, each with 100% flow capacity, are operable at any time when heat removal from the primary system is through the steam generators. When two independent 100% capacity flow paths are not available, the capacity shall be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the plant shall be placed in a cooling mode which does not rely on steam generators for cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
When at least one 100% capacity flow path is not available, the reactor shall be made subcritical within one hour and the facility placed in a shutdown cooling made which does not rely on steam gen-erators for cooling within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or at the maximum safe shutdown rate."
The licensee has stated in Reference 2 to this evaluation that five separate surveillance procedures presently in use at Rancho Seco assure that two inde-pendent auxiliary feedwater (AFW) flow paths, each with 100% flow capacity, are operable at any time in order to meet the requirements of its Technical Specifications.
The licensee stated that the Rancho Seco Technical Specifi-cation requirements meet or exceed the stated times for each mode of operation addressed in item 8 of the bulletin.
The staff noted, however, that the Rancho Seco Technical Specifications (TS) do not contain the requirement that the reactor be made subcritical within one hour wnen auxiliary feedwater is not available. Furthennore, the TS require only that a hot shutdown procedure De completed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> wnen auxiliary feedwater is not available.
Since the one hour require-ment is not presant, and hot shutdown (reactor coolant temperature greater O
1601 207
. than or equal to 525'F) is not "a cooling mode which does not rely on '
stean generators for cooling," the Rancho Seco TS do not meet the require-ments of the bulletin. The staff has discussed these discrepancies with the licensee, and a resolution is being pursued in connection with a review of all TS changes necessary as a result of Bulletins79-05A and 79-05B and the Commission Order of May 7,1979. These TS changes will be the subject of a separate safety evaluation which is being developed to support the necessary license amendment.
The staff also noted that bin normally open, motor-operated cross-tic valves are present in the AFW pump discharge lines.
In the event of a single passive failure, such as a pipe rupture in one of the discharge lines, both feedwater paths could be rendered inoperable.
In a telephone conversation on April 18, 1979, between the staff and the licensee, we requested that the licensee address this issue.
In a letter dated Aoril 19, 1979 (Reference 3),
the licensee stated that the two AFW pumps are powered from separate power supplies and that the motor-operated cross-tie valves are powered from separate Class IE power supplies. The licensee considers that this arrangement provides the necessary independence, and assures flow to the steam generators in the event that only one AFW pump starts, even though operator action would be required to shut the valves in the event of a break in one of the AFW pump discharge lines. These valves can be operated from the control room and the operator can verify flow to the steam generators by observing flow rate indica-tion, installed as part of the short-term requirements of the Commission Order of May 17, 1979.
The staff concurs with the licensee's justification for keeping the cross-tie valves open during normal operation.
The NRC staff finds the licensee's responses to item 8 of IE Bulletin 79-05A complete.
O 16M 208 Item 9:
" Review your operating modes and procedures for all systems designed to transfer potentially radioactive gases and liquids out of the primary containment to assure that undesired pumping of radioactive liquids and gases will not occur inadvertently.
In particular, ensure that such an occurrence would not be caused by the resetting of engineered safety features instrumentation.
List all such systems and indicate:
a.
whether interlocks exist to prevent transfer when high radiation indication exists; and, b.
whether such systems are isolated by the containment isolation signal."
The licensee documented its response to this item in Reference 2 to this evaluation.
The licensee stated that for all systems that could transfer potentially radioactive liquids and gases out of the primary containment, a safety features actuation signal on low reactor coolant system pressure or high reactor building pressure, would close all valves which could cause this transfer.
The safety features isolation for Rancho Seco can only be overridden by placing the appropriate controls on the safety features actuation systems (SFAS) panels in " manual" and depressing the "open" push buttons as desired.
The licensee further stated that the operator is cautioned in procedures to consult the Technical Specifications prior to placing any portion of the SFAS cut of service. When the safety features actuation signal clears, all valves must be manually repositioned to their normal position.
Finally, the licensee stated that if a containment purge was in progress, the purge would be termi-nated by the SFAS, or by a high radiation signal from the reactor building monitor.
Either of these signals will secure the containment supply and exhaust fans.
O 160( 209 i
. The subject of the containment isolation valves being opened to allow purging during normal operation is presently under staff review.
In our letter of November 28, 1978, we requested the licensee to provide justification for continued purging at Rancho Seco and pending NRC staff review of its justi-fication, to limit purging to an absolute minimum of 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year.
The licensee provided a justification for purging in its letter of January 4, 1979. The licensee also agreed in a letter dated June 15, 1979, to limit purging during power operation to 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year.
This matter is presently under review by the NRC staff.
The NRC staff finds that the licensee has been responsive to item 9 of IE Bulletin 79-05A and that further resolution of purging during power operation will be handled under the NRR Generic Issues Program. Therefore, the NRC staff considers the licensee's response to this item complete.
Item 10:
" Review and modify as necessary your maintenance and test procedures to ensure that they require:
9 a.
verification, by inspection, of the operability of redundant safety-related systems prior to the removal cf any safety-related system from service:
b.
verification of the operability of all safety-related systems when they are returned to service following maintenance or testing; and, c.
a means of notifying involved reactor operating personnel whenever a safety-related system is removed from and returned to service."
The licensee's response to item 10 of this bulletin is documented in Reference 2 to this evaluation.
With regard to sucparagrach a above, the licensee stated tnat an Outage Coordinator.
who maintains a Senior Reactor Ocerator's License, reviews any work and properiy indicates the logging of tests for redundant safety-related equipment as
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1607 210 required in the Technical Specifications (TS).
The TS for Rancho Seco require the testing of the redundant component (s) prior to maintenance on a component (s) in a safety-related system.
In addition, the TS list the safety-related systems for which this requirement is applicable. We have also reviewed the Rancho Seco surveillance procedures and found that this same requirement holds for the AFW system.
With regard to subparagraph b above, the licensee stated that the operability of a safety related system, following maintenance, is accomplished by perform-ing a specified post-maintenance test determined by the cognizant engineer.
The results of this test are then evaluated by the cognizant engineer and sent to the shift supervisor for concurrence, provided the acceptance criteria are met.
In response to the Commission's Order of May 7, 1979, the licensee has revised its procedures to require two operators to perform and verify, by signature, correct valve alignment of the AFW system following maintenance.
With regard to subparagraph c, the licensee stated that the " work request" procedure requires shift supervisor notification prior to any work on safety-related equipment, when the system may be removed from service.
The procedure also calls for the signature of the shift supervisor when returning the system to operation.
The shift supervisor is required to state system conditions in the Shift Supervisor's Log.
The staff was concerned that control room operators may not have explicit information about the status of safety-related systems.
In response to this concern, the licensee stated that maintenance procedures require tagging out safety systems on the operating panels in the control room.
This ensures that all on-duty operators know when systems are in or out of service.
The NRC staff finds the 'icensee's responses to item 10 of IE Bulletin 79-0SA complete.
Item 11:
"All operating and maintenance personnel should be made aware of the extreme seriousness and consequences of the simultaneous blocking of coth auxiliary feecwater trains at the Three Mile Island Unit 2 plant anc ather actions taken during the early phases of the accident."
u 16W 211
(
. The licensee has documented in Reference 2 to this evaluation that all operating personnel have had training on this matter and are aware of the extreme serious-ness and consequences of the THI-2 accident.
The licensee further stated that maintenance personnel do not change valve positions at Rancho Seco and have been instructed that they do not have this authority.
The staff considered that the licensee's response wou'd be acceptable, provided that the instructions given to maintenance personnel stressed the reasons why they do not have this authority an_d the potential consequences of incorrectly assuming this authority.
The licensee has confirmed that this action has been taken.
The NRC staff finds the licensee's responses to item 11 of IE Bulletin 79-05A complete.
Item 12:
" Review your prompt reporting procedures for NRC notification to assure very early notification of serious events."
The licensee has documented its responses to this item in Reference 2 to this evaluation.
The response from the licensee outlines the procedural controls that have been established for NRC notification of serious events.
Reference is made by the licensee to the guidance and requirements established in Regu-latory Guide 1.16 (" Reporting of Operating Information-Appendix A Technical Specifications") for reportable occurrences and in 10 CFR 20.403(a) for radio-logical incidents.
In addition for cases of overexposure, fire, sabotage or plant evacuation the Rancho Seco Emergency Plan Implementing Procedures require NRC notification.
The NRC staff finds the licensee's response to item 12 of IE Bulletin 79-05A complete; however, IE Bulletin 79-05B expands tne licensee's responsibility in this area.
Further discussion of this matter can be found under the staff's evaluation of item 6 of IE Bulletin 79-058.
EVALUATION OF RESPONSES TO IE BULLETIN 79-052 Item 1:
" Develop procecures and train oceration personnel on methods of establishing and maintaining natural circulation.
The procedures and training must include means of monitoring heat removal efficiency
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16M 212 by available plant instrumentation.
The procedures must also contain a method of assuring that the primary coolant system is subcooled by at least 50*F before natural circulation is initiated.
In the event that these instructions incorporate anticipatory filling of the OTSG prior to securing the reactor coolant pumps, a detailed analysis should be done to provide guidance as to the expected system response.
The instructions should include the following precautions:
a.
maintain pressurizer level sufficient to prevent loss of level indication in the pressurizer; b.
assure availability of adequate capacity of pressurizer heaters, for pressure control and maintain primary system pressure to satisfy the subcooling criterion for natural circulation; and c.
maintain pressure /tereperature envelope within Appendix G limits for vessel integrity.
Procedures and training shall also be provided to maintain core cocoing in the event both main feedwater and auxiliary feedwater are lost while in the natural circulation mode."
The licensee has documented its response to this item in References 5, 6 and 7 to this evaluation.
Reference 5 stated that plant operating procedure B.4
(" Plant Shutdown and Cooldown") was being revised to include methods of estab-lishing and maintaining natural circulation including the monitoring of heat removal efficiency and subcooling criteria described in the bulletin.
- However, the licensee stated that anticipatory filling of the steam generator was not expected to be necessary but tnat B&W was in the process of performing a review of this area.
In addition, the licensee stated the procedure had been revised to reflect the proper operator response upon loss of both main and ll 16a:7 213
. auxiliary feedwater flow whilo in the natural circulation mode.
In Reference 6, the licensee reported that the results of the B&W analysis on the advisability of anticipatory filling of the steam generators prior to securing the reactor coolant pumps showed this action to be beneficial. A copy of this analysis was attached to Reference 6.
As a result of this analysis, the licensee stated that it had modified operating procedures to incorporate the results of the analysis including precautions for proper pressurizer level and heater capacity.
Precautions were also added to ensure that the operator maintains the proper pressure / temperature relationship to remain within acceptable regions of the Rancho Seco Technical Specifications.
Reference 7 forwarded a revised Figure 1 originally sent as Enclosure 2 to Reference 6.
The Office of Inspection and Enforcement (IE) reported in Reference 12 that the above items had been verified to be complete and that proper operator training had been conducted on the revised procedures.
The NRC staff finds the licensee's response to item 1 of IE Bulletin 79-05B complete.
Item 2:
" Modify the actions required in Item 4a of IE Bulletin 79-05A to take into account vessel integrity considerations:"
'4.
Review the action directed by the operating procedures and training instructions to ensure that:
a.
Ope:ators do not override automatic action of engineered safety features, unless continued coeration of encineered safety features will result in unsafe olant conditions.
For examole. if continued coeration of encineered safety features would threaten reactor vessel intecrity the HPI 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 has been automatically actuated because of low pressure condition, it must remain in operation until either:
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16M 214
. (1) Both low pressure injection (LPI) pumps are in operation and flowing at a rate in excess of 1000 gpm each and the situation has been stable for 20 minutes; or, (2) The HPI system has been in operation for 20 minutes, and 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, the HPI shall be reactivated.
The degree of subcooling beyond 50 degrees F and the lenath of time HPI is in coera-tion shall be limited by the cressure/ temperature considerations for vessel integrity. '"
- [ NOTE:
Underlined portions are modifications to, and supersede, IE Bulletin 79-05A]
The licensee's reply to this item is documented in Reference 5 to this evaluation.
In this response the licensee stated that the Rancho Seco emergency procedures had been modified to reflect the requirements of this item.
However, suosequent to the issuance of IE Bulletin 79-05B, further refining of the HPI termination criteria took place based on guidelines developed by B&W and reviewed by the NRC staff.
The licensee has developed these guidelines into plant specific emergency procedures for the Rancho Seco facility.
The present HPI termination criteria, as defined in Revision 13 to Emergency Procedure 0.5 (" Loss of Reactor Coolant / Reactor Coolant System Pressure) are as follows:
"With HPI automatically actuated due to low RCS pressure, 00 NOT terminate SFAS flow until:
1)
Both low pressure injection (LPI) pumps are in operation and flowing at a rate in excess of 1000 gpm each and the situation has been stacle for 20 minutes;
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16B5 215 OR:
2)
All hot and cold leg temperatures are at least 50* belaw the saturation temperature for the existing RrS pressure and the hot leg temperatures are not more than 50 hotter than the secondary side saturation temperature.
If 50* subcooling cannot be maintained, tr.? HPI shall be reactivated.
RCS relief valve operations will prevent RCS pressure from exceeding 2750 psig.
(RCS pressure recorder will indicate pressure instability
- 2500 psig).
Maintaining 50" subcooling will ensure pressure / temperature limits are not exceeded until RCS pressure reaches 800 psig.
Further pressure reduction. requires use of Figure 101-2a of the Process Standards."
The staff finds the licensee's response to item 2 of IE Bulletin 79-058 complete.
Item 3:
"Following detailed analysis, describe the modifications to design and procedures which you have implemented to assure the reduction of the likelihood of automatic actuation of the pressurizer PORV during anticipated transients.
This analysis shall include consideration of a modification of the high pressure scram setpoint and the PORV opening setpoint such that reactor scram will preclude opening of the PORV for the spectrum of anticipated transients discussed by B&W in Enclosure 1.
Changes developed by this analysis shall not result in increased frequency of pressurizer safety valve operation for these anticipated transients."
The licensee documented its response to this item of the bulletin in References 4 and 5 to this evaluation.
The licensee stated that it had reviewed the design aspects of the reactor protection system and operating procedures which could affect automatic actuation of the pressuri:er pilot operated relief valve.
The review included the analysis of anticipated transients performed by B&W.
The licensee's analysis concluded that lowering the high pressure reactor trip setcoint from 2355 psig to 2300 psig, and raising the pilot operated relief valve setcoint from 2255 psig to 2450 psig proviced the requested reduction in automatic PORV actuation during anticipated transients.
Changes were inccrporated
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160f 216 into the appropriate plant operating procedures.
In addition, the licensee stated that the setpoint changes had been initiated.
In a follow-up letter, Reference 5, the licensee stated that the requirements of item 3 had been completed, within the 24-hour time period required by the bulletin.
Verifica-tion of setpoint changes was completed by IE and reported in Reference 12.
The NRC staff finds the licensee's response to item 3 of IE Bulletin 79-058 complete.
Item 4:
" Provide procedures and training to operating personnel for a prompt manual trip of the reactor for transients that result in a pressure increase in the reactor coolant system. These transients include:
a.
loss of main feedwater; b.
turbine trip; c.
main steam isolation valve closure; d.
low OTSG level; and, f.
low pressurizer level."
The licensee has documented its response to this item in Reference 5 to this evaluation. The licensee stated that procedures were modified to require prompt manual trip of the reactor upon:
loss of main feedwater, turbine trip, loss of offsite power, low steam generator water level and low pressurizer level.
Since the Rancho Seco design does not include main steam isolation valves, item 4.c of this bulletin is not considered applicable.
In reviewing the requirements of the bulletin, it is noted that this item lists low pressuri:er level as an example of " transients that result in a pressure increase in the reactor coolant system." However, a low pressuri:er level may result from low reactor coolant system pressure, and a prompt manual reactor trip in some cases may not neuessarily be the most advantageous action for the operator to perform (i.e., An overcooling event will cause both pres-suri:er level and RCS pressure to decrease.
In this case, a reactor trip
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' 6M 217
. would cause their parameters to decrease further).
Therefore, staff has reviewed the two procedures (Loss of Coolant / Loss of. Pressure and Loss of Makeup / Letdown) cited by the licensee as having been revised to require a manual reactor trip on low pressurizer level.
The staff considers, that in the context of the complete set of actions required by these revised procedures, a manual reactor trip is appropriate.
Item 5 of this bulletin discusses the status of providing an automatic reactor trip for certain of these transients.
The NRC staff finds the licensee's response to item 4 of IE Bulletin 79-058 complete.
Item 5:
" Provide for NRC approval a design review and schedule for implemen-tation of a safety grade, automatic anticipatory reactor scram for loss of feedwater, turbine trip, or significant reduction in steam generator level."
In Reference 8 to this evaluation, the licensee provided simplified drawings and a schedule for installing a safety grade, automatic anticipatory reactor trip for loss of main feedwater and turbine trip.
Based on an analysis performed by B&W, the licensee stated that it did not feel that a low steam generator level trip would serve as an anticipatory trip since the reactor would normally trip on high reactor coolant system pressure prior to tripping on low steam generator level. With regard to the schedule, the licensee stated that procure-ment of equipment would take approximately 9 months following NRC approval of the design.
Installation of the modification would occur during the first refucring outage following receipt of the required equipment.
Subsequent to the issuance of IE Bulletin 79-05B, the Commission Order of May 7, 1979 was issued to the licensee.
One of the immediate actions required of the " censee, based on this Order, was to install a control grade reactor trip for loss of main feecwater and turbine tr;p.
The Order requires that for continued long-term operation, the licensee must upgrade this circuitry to meet safety grade criteria. A letter was issued to the licensee, dated Septemoer 7,1979 (Reference 14), which fomarded a request for additional 0
16M 2\\8 information on the proposed design. This information is needed before the staff can approve the proposed design for Rancho Seco.
In addition, this letter requested that the licensee expedite it:, installation schedule such that installation and testing could be completed within about 6 months following NRC staff approval of the design.
The NRC staff finds that the licensee has been responsive to item 5 of IE Bulletin 79-05B and that any further follow-up action on this matter will be handled under the long-term portion of the Commission Order of May 7, 1979.
Therefore, the NRC staff finds the licensee's response to this item complete.
"The actions required in item 12 of IE Bulletin 79-05A are modified Item 6:
as follows:
Review your prompt reporting procedures for NRC notification to assure that NRC is notified within one hour of the time the reactor is not in a controlled or expected condition of operation.
Further, at that time an open and continuous comunication channel shall be established and maintained with NRC."
The licensee's response to this iten is contained in Reference 5 to this evaluation. This response was further clarifi ed in Reference 1,6.
In its response, the licensee committed to informing the NRC within one hour from the time the reactor is not in a controlled or expected condition of operation.
The licensee has defined the NRC as the Duty Officer at the NRC Operations Center in Bethesda, Maryland. The licensee has directed the Rancho Seco operators to notify the NRC within one hour if either of the following two condi tions are met: (1) if an unscheduled change of more than 50 MWe in electrical generator output occurs or (2) if the reactor is shutdown and In addition to the an unplanned cessation of reactor coolant flow occurs.
above requirements, any time the reactor is in an off normal transi ent ccn-dition, the Shift S;pervisor will notify one of four designated management officials : (1) Manager of Nuclear Operations, (2) Plant Superintendent, (3)
Engineering and Nality Control Supervisor or (4) Operations Supervisor. These designated individuals nold Senior Reactor Operators Licenses. Based ugen ne management official's evaluation of the situation, he will determine NRC reoart-ability and :ake the necessary actions, one of which will be having the reporting Shift Supervisor use the dedicated telephone to notify tne NRC of the situation.
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16M 219 The licensee has installed a speaker phone in the control room, which is connected to the NRC Operations Center. The spesker phone will allow the control room oper-ators to keep the NRC informed of any actions taking place to cope with a plant emergency. The open comunications channel shall remain open for each notification until the NRC personnel agree that it is no longer required.
The NRC staff finds the licensee's response to item 6 of IE Bulletin 79-05B compiete.
Item 7:
" Propose changes, as required, to those technical specifications which must be modified as a result of your implementing the above items."
The licensee's response to this item is contained in Reference 8 to this eval-uation. The licensee considered that the Technical Specification (TS) pro-viding for a high pressure reactor trip at 2355 psig was the only one that could possibly be changed as a result of implementing the requirements of IE Bulletin 79-058.
However, it stated that the lowered setpoint of 2300 psig is within the present TS requirements (since 2355 psig is stated as a maximum value in the TS), and proposed that no changes be made at that time.
The licensee did state that additions to the TS resulting from the implementation of the reactor trips required by item 5 of this Bulletin would be appropriate when the plant modifications were made.
By Reference 12 to this evaluation, which forwarded the staff's evaluation of tne licensee's compliance with the Commission Order of May 7, 1979, the licensee was requested to provide the staff with TS changes that reflected the addition of the anticipatory reactor trip and changes to the setpoints for the PORV and the high pressure reactor trip.
These TS changes were forwarded to the NRC for review by Reference 9 to this evaluation.
These proposed TS changes are presently under review and will be the subject of a separate safety evaluation wnich is being developed to support the necessary license amendment.
The NRC staff finds the licensee's response to item 7 of IE Sulletin 79-058 complete.
O 16g 220
SUMMARY
OF OUTSTANDING ITEMS As a result of the staff's review of the licensee's responses to IE Bulletins79-05A and 79-058, the staff has identified gertain items for which additional information must be obtained in order to resolve these matters.
The list below summarizes these matters.
A more detailed discussion of these items is provided under the appropriate IE Bulletin item of this evaluation.
IE SULLETIN 79-05A Item 1:
The staff believes that a temperature monitor alone, downstream of the PORV is not always a valid indication of PORV position, and that a more direct means of monitoring the PORV position should be avail-able.
This item will be resolved as part of the licensee's compliance with Section 2.1.3.a of NUREG-0578 (Section Title
" Direct Indication of Power-Operated Relief Valve and Safety Valve Position for PWRs and BWRs").
NUREG-0578 requires implementation of this item by January 1, 1980.
Therefore, no additional action will be taken on this matter under IE Bulletin 79-05A.
Item 4:
Ihm 4.c required certain actions be taken by the licensee with respect to operation of the reactor coolant pumps following HF7 initiation.
The requirements of IE Bulletin 79-05C supersede the requirements of item 4.c of this bulletin.
Responses to IE Bulletin 79-05C are presently under review by the staff.
Following completion of this review, the staff will publish a separate evalua-tion covering the matter for all PWRs.
This evaluation is scheduled to be publisned later this year as a NUREG document.
Therefore, no additional action will be taken on this matter under IE Bulletin 79-05A.
Item 5:
Upon automatic initiation of safety injection, all lines whose isolation does not degrade core cooling capability are isolated for the Rancno Seco containment with the exception of the lines necessary to assure continued operation of the RCPs and control rods. The 16k221 acceptability of not isolating these lines will be reviewed as part of the NRC staff's evaluation of the licensee's compliance with Section 2.1.4 of NUREG-0578 (Section Title
" Diverse and More Selective Containment Isolation Provisions for PWRs and BWRs").
Therefore, no additional action will be taken on this matter under IE Bulletin 79-05A.
Item 8: A separate safety evaluation to support the required license amendment will be prepared that will include all Technical Specification changes necessary as a result of the licensee's implementation of IE Bulletins79-05A and 79-05B and the t.onnission Order of May 7, 1979. Therefore, no additional action will be taken on this matter under IE Bulletin 79-05A.
Item 9:
The subject of containment purging during power operation is presently under staff review as part of the NRR Generic Issues Program (Task - B-24:
" Venting and Purging of Containment While at Power Operation and Effects on LOCA").
Any additional information needed to resolve this matter will be developed under this generic activities task.
Therefore, no additional action will be taken on this matter unGer IE Bulletin 79-05A.
IE BULLETIN 79-05B Item 5:
This item required the licensee to submit a proposed design and schedule for a safety grade, automatic anticipatory reactor trip.
The Commission Order of May 7, 1979, requires that this feature be installed as part of the long-term requirements of the Order.
Therefore, any additional information required of the licensee in this matter will be reviewed as part of the NRC staff's evaluation of the licensee's compliance with long-term portion of the Order and no additional action will be taken under IE Bulletin 79-05B.
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16M 222 Item 7:
A separate evaluation to sur, port the required license amendment will be prepared that will include the licensee's proposed modifications to the Technical Specifications.
Therefore, no additional action will be taken on this mattar under IE Bulletin 79-05B.
CONCLUSIONS Based on our review of the information provided by the licensee ia response to IE Bulletins79-05A and 79-05B, and with the exception of the outstanding items identified above, we conclude that the licensee has acceptably responded to these Bulletins.
The actions taken by the licensee demonstrate its under-standing of the concerns and implications of the TMI-2 accident as they relate to the Rancho Seco Nuclear Generating Station.
These actions have resulted in added assurance for the continued protection of the public health and safety during plant operation.
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160f 223 REFERENCES 1.
Letter from J. J. Mattimoe (SMUD) to R. H. Engelken (NRC), dated April 11, 1979, providing responses to items 1, 2, 3, 4a and 5 of IE Bulletin 79-05A.
2.
Letter from J. J. Mattimoe (SMUD) to R. H. Engelken (NRC), dated April 16, 1979, providing responses to items 4b through 4d, and 6 through 12 of IE Bulletin 79-05A.
3.
Letter from J. J. Mattimoe (SMUD) to R. H. Engelken (NRC), dated April 19, 1979, providing clarification of response to item 8 of IE Bulletin 79-05A.
4.
Letter from J. J. Mattimoe (SMUD) to R. H. Engelken (NRC), dated April 22, 1979, providing response to item 3 of IE Bulletin 79-053.
5.
Letter from J. J. Mattimoe (SMUD) to R. H. Engelken (NRC), dated May 2, 1979, providing responses to items 1, 2, 4 and 6 of IE Bulletin 79-05B and providing documentation that item 3 of IE Bulletin 79-05B was completed within the 24-hour requirement of the Bulletin.
6.
Letter from J. J. Mattimoe (SMUD) to R. H. Engelken (NRC), dated May 11, 1979, providing additional response to item 1 of IE Bulletin 79-05B and forwarding " Natural Circulation - Intentional Securing of Reactor Coolant Pumps."
7.
Letter " rom J. J. Mattimoe (SMUD) to R. H. Engelken (NRC), dated May 14, 1979, providing corrected Figure 1 to the analysis provided in Reference 6 above.
8..
Letter from J. J. Mattimoe (SMUD) to R. H. Engelken (NRC), dated May 21, 1979, providing responses i.o items 5 and 7 of IE Bulletin 79-053.
9.
Letter from J. J. Mattimae (SMUD) to R. W. Reid (NRC), dated July 2,1979, providing revised Technical Specifications for modifications completed in compliance with the Order of May 7, 1979.
10.
Letter from J. J. Mattimoe (SMUD) to R. H. Engelken (NRC), dated March 31, 1978, foNarding Reportable Occurrence - R0-78-01 for Rancho Seco.
11.
Letter from J. J. Mattimoe (SMUD) to R. H. Engelken (NRC), dated January 26, 1979, forwarding Reportable Occurrence - R0-79-01 for Rancho Seco.
12.
Letter from H. R. Denton (NRC) to J. J. Mattimoe (SMUD), dated June 27, 1979, permitting resumption of operation in accordance with the terms of the Order of May 7, 1979 and enclosing the " Evaluation of Licensee's Comoliance with the NRC Order Dated May 7,1979 - Sacramento Municipal Utility District - Rancho Seca Nuclear Generating Station - Docket No. 50-312."
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16M 224
13.
NUREG-0578, "THI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," July 1979.
14.
Letter from R. W. Reid (NRC) to All Babcock & Wilcox Operating Plants, dated September 27, 1979, requesting additional information concerning the upgrade of the anticipatory reactor trip (loss of feedwater and turbine trip).
15.
Letter from J. H. Taylor (B&W) to D. F. Ross (NRC), dated August 17, 1979, forwarding B&W's generic report BAW-1564 entitled " Integrated Control System Reliability Analysis."
- 16. Letter from R. J. Rodriguez (SMUD) to R. H. Engelken (NRC), dated October 26, 1979, revising response to item 12 of IE Bulletin 79-05A and item 6 of IE Bulletin 79-05B.
O 16W 225
.