ML19269F474

Evaluation of Licensees Responses to IE Bulletins79-05A & 79-05B

ML19269F474
Person / Time
Site:	Rancho Seco
Issue date:	11/23/1979
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML19269F471	List: ML19269F470 ML19269F474
References
IEB-79-05A, IEB-79-05B, IEB-79-5A, IEB-79-5B, NUDOCS 7912210078
Download: ML19269F474 (29)

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Of<f EVALUATION OF LICENSEE'S RESPONSES x.;g.J s TO IE BULLETINS79-05A AND 79-05B SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GEhERATING STATION DOCKET NO. 50-312 2167 004

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. INTRODUCTION On March 28, 1979 the Three Mile Island Huclear Power Plant, Unit 2 (TMI-2)

. experienced core damage which resulted frem 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 This bulletin was superseded and expanded by IE Bulletin 79-05A dated April 5,1979, and by IE Bulletin 79-058 dated April 21, 1979.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 79-0.58, 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 Bulletin 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 Comission issued an Order dated May 7,1979, which confirmed the licensee's commitment to make "certain modifications to plant equipment and procedures, an'd'to complete specific ~ operator training-and analyses of plant behavior.

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-Due-to the--overlap-in-the req'uirements of the Order and the bulletins, the Order is referenced several times in this evaluation.

In addition, the NRC's L'essons Learned Task Force has recently completed its report (NUREG-0578)

_ detailing slioit-term recomendations that are to be implemented for all

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2-operating reactor plants in light P tae accident at TMI-2. This report has also been referenced in this evaluation. The NRC staff's evaluation of the licensee's compliance with the short-tem portion of the Commission Orcar was issued on June 27,1979(_ Reference 121, Separate evaluations of the lice 5see's compliance with the png-tem portion of the Comnission Order and MUREG-0573 will be issued at a future date.

The NRC staff's evaluation of tha 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 po: tion 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 additic i to the review of circumstances descriced 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 reviewed Enclosure 1 to IE Bulletin 79-05 and the preliminary sequence of events enclos~ed with IE Bulletin 79-05A. The licensee, assisted by B&W, assessed ths adetitiacy 'o~f Raricho Seco.to safe.ly, sustain 'transte'nts :uch -

as the one _whichoccIrrei at TMI-2.1-fts' revie'w-ider)tified ttie iase~sil human,-

design animec~nanical faf-lures which resulted in. the-core; damage and r.adiation

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releases. at TMI-2, as -are described in the " Description of Circumst.ances" portion of IE Bulletin 79-05A.

Details _of this review are documented in-

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Reference 1 to.this evaluation'.

The staff.has 'reviesed'this documentland

-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 elactromatic relief valve (PORV), which opened during the initial pressure surge, failed to close when the pressure decreased below the actuation level.

In Referencs 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 i.he PORV position should be available. This item will be.

resolved as part of the implementation of Section 2.1.3a of NOREG-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 high1 radioactive water from the PORV discharge was pcmped 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 i;olation would isolate the Rancho Secc 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 o_f radioactive liquids and gases-outLof the: containment will not occur.. inadvertently._:The issue. of_ containment isolation.is addressed. in our eva'1uatio'n of the respbnse'to-item 9; -

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The description-of circumstances in--IE Bulletin;79-05A discusses-the adverse effect of ~ intermittent operation orthe:HPr system it TMI-2; The lic6nsee stated in Reference 1 that recent revisioris to the operating and emergency ~

procedures at Rancho Seco, as required by items 4a, b and d of IE Bulletin 79-05A, will preclude occurrence of a similar event at Rancho Seco.

This matter is addressed in our evaluation of the responses to item 4.

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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.

As discussed in References 1 and 2, the licensee had modified its pro.cedures to assure at least one RCP per loop remained operating during a loss of reactor coolant / loss of reactor coolant systern 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

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suparseded 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 direct 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:

" Review any transients similar to the Davis-Besse event (Enclosure 2 of IE Bulletin 79-05) and any others which contain similar elements

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from the enclosed chronology (Enclosure 1) which have occurred at

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your facility (ies).

If any significart deviations from expected performatice are identified in your revi,ew, provide details and an analysis of tne safety significance together with a description of any corrective. actions taken.

Reference may be made to previous

_information p_rovided_ to the NRC, if appropriate, in responding..to. _

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'In response to itee 2 o7-Bu'let'in 79-05A, the licensee stated in Reference T

._that.it had reviewed transients which had occurred at Rancho Secc..to determine if any. had elements-siai.lar to the. chronology-of' events at -TMI-2 (March-Ea,'

1979)'and Davis-Besse l_(November' 29, 1977)., Based upon this, review,.the ~

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licensee stated tr.at it had not identified any transients which were similar; however, it had reviewed one transient in which the cooldown resulted in operation outside the Technical Specification pressure / temperature limits.

. 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 developsd an emergency procedure that allows plant condiMons to be stabilized following a loss of NN1 pcwcr.

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,

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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 desirable 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 failu*e modes and effects analysis of the ICS. This report was su,bmitted 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 resnonse to item 2 of IE Bulletin 79-05A complete.

Item 3:

" Review the actions required by your operating procedures for coping

(~with jransients~ and accidents, with particular attention to:

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..- am recognitien.cf__the".possib,il-ity,cf_?for.?ing_yoids...in..the primary coolant system large.enough.to compromise._the. core c.co. ling capability; especially natural circulation capab_il,ity;. _,

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-operator action required to prevent the formation of such voids; and 9

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operator action required to enhance core cooling in the event such voids are formed."

As a result of the TMI-2 accident, all licensed operators have received additional training to enable t' am to cope with transients and accidents.

The specific ttaining received to comply with subparagraphs a, b and c above are documented in Reference 1 to this evaluation.

Specific staff cmmnts 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, wnich would demonstrate the THI-2 event and actively engage them in the analyzing and correcting abnormal transient situations including steau 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 fr ~

all licensed operators was required by the Ccmmission's Order of May 7,1979

_and this training was completed for al1 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 eva1ua' tion, IE Bullet ~in 79-05C requires licensees ~to immidiately

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-trip al1- ~ operating redEtdr tooTanfWmps i'n -ths event af Freactor trip.and.

initiation of HPI caused by reactor coolant system low pressure.

Rancho Seco operating procedures have~ been todified to reflect the requirements of ~It-Bulletin 79-05C.

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Additional requirements in the area emergency procedures for transients and accidents have been recommended in-Section 2.1.9 of NUREG-0578.

To comply with these requirements, the licensee is actively engaged in developing operater guidelines which cover inadequate core cooling and other abnormal transients.

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. 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 omplete.

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 infection (HPI) system has been automatic-ally actuated because of low pressure condition, it must remain in operation until either; (1) both low pressure infestion (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 pressur+

If the 50 degree subcooling cannot be maintained

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. operating precedures. currently,-or are revised to,-specify-that

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in the event of HPI initiation, with reacto'r coolant pumps (RCP) operating, at least one RCP per loop shall remain cperating.

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operators are provided additional __information and i.nstructions to not rely upon pressurizer level indication alone, but to

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. also examine pressurizer pressur other plant parameter indications in evaluating plant conditions, e.g., water inventory in the reactor primary system."

IE Bulletin 79-053 modified the actions required in subparagraphs a and b above to take into account riressure vessel integrity considerations.

Evaluation of this matter is discussed unu.c itcm 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 79-05C.

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 0perator 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 subcoo!ed. 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 Ji_cer$s'ees up[ grade react 6r-instrumestati[6:i _to' provide the o~peritor, with an

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unambiguous -indicatio.n of vessel water level and core cooling-adequacy.

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.3 The NRC staff finds that -the licensee has been r.es;5cnsive to item 4.cf,IE Bulletirr-79-0EA-and that.a'ny -futther follow up action on item 4.c will be.

handled under the NRC staff review of IE Bulletin 79-05C.

Therefore, the NRC staff considers the licensee's respo~n'se to this item complete.

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Itom 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 requirements to assure that valves are pesitioned (open or closed) it. a manner to ensure the proper operation of engineered safety features. Also review related procedures, such as those for 'aintenance and testing, to ensure that such valves are m

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 Ccamissions's Orde; of May 7,1979.

A revi, 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."

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~The licensee:has docu:gented in Reference.2 to-this evaluathn, that contain-

~ ment ~ isolation ~at Rancho-Seco occurs at either 1,600 psig reactor coolant ~'

- system pressure-or at-4-psig reactor containment oressure.

The licensee has~

. stated that_it has reviewed containment; isolation design and procedure.s and n

.has-found them-acceptaMe...Al.1 Jines notgequired for: safety features are.

isolated upon initiation of containment isolation except-~ lines that are necessary to assure continued operation of the reactor coolant pumps 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

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with Section 2.1.4 of NUREG-0578.

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The NRC staff finds that the licensee has been responsive to item 6 of IE Bulletin 79-05A and that any further resciution 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 correct 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 Yalve 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 specii'ically addressed in the bulletin, the staff was concerned a' bout the ' operation anifr~eliabil'ity of the[tho pne'umaticall[opeYatid ficw I

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control valves in the Rancho Seco' aux'iliary feedWater: system.

These value.s are controlled by the 103- - As part-of complying with the immediate acti ns of -

the-Commission's Order of May 7,1979, the licensee-developed a, procedure that allows cperator control of the auxiliary feedw'ater flow independent of the ICS through the-safety grade bypass valves.

This action h n resolved the staff's r--

concern about tne pneusatically-cpirated Valke's.~~ ~

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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 op :.able 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 made 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 reacto'r shall be made subcritical within one hour and the facility placed in a shutdown cooling mode 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 survefilance Tracedures presently in' use at Rancho Seco assure that two inde-pendent auxiliary feedwater (AFW) flow paths, each with 100%. flow capacity, are operabia at any time in order to mee, the requirements of its Technical r

Specifications. The licensee stated that the Rancho Seco Technical Specif1-cation requirements meet or exceed the stated times for each mode of operation addressed in item 8 of the bulletin.

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The staff ~noted, however,that 'ths"Ran-h'o Se'co' Techni~ cal' Specif~ications (TS) do not contain the_ requirement _that the reactort be made subcritical-withi.rn one

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hour when auxiliary feedwater is not available.__Eurthemore,z the..TS_-

require only-that a hot shutdown procedure be completed within 12-hours

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wheiiTausiTiary~~feedwatht.is not_available. Since the ~one~ hour requiYe

ment is not present, and hot shutdown (reactor coolant temperature greater

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. than or equal t.' 525'F) is not "a cooling made which does not rely or-stean generators for cooling," the Rancho Seco T5 do not meet the requirt-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 79-053 and the Commission Order of May 7,1979. These TS changes will be the subject of a separate safety evaluation which is being dsveloped to support the nece>sary license amendment.

The staff also noted that tuo normally open, motor-operated cross-tie valves are present in the AFW prep 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.

I.. a telephone conversation on April 18, 1979, between the staff and the licensee, we requested that the licensee address this issue.

In a Tetter dated April 19, 1979 (Reference 3),

the licensee stated that the two AFW pumps are powered from separate power supplies and that the coto operated cross-tie valves are powered from separate Class IE power supplies. The licensee considers that this arrangement provides the 1ecessary independence, and assures flow to the steam generators in the event that only one AFW pump starts, even though operator action would be reouired 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~st'affconcuriEiththelicensee'sjustification~for

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keeping the c~ross-tie. valves open during normal operation.

n The~NRC staff finds the licensee's~ responses to item 8 of IE Bulletin 79-05A

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. Item 9:

" Review your operating modes and procedures for all systems designed to transfer potentially radioactive gases and liquids or of the primary containment te assure that undesired pumping of radioactiva liquids and gases will not occur inadvertently.

In particular, ensure that such an occrrrence would not be caused by the resetting of engineere~d 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 thi~s evaluation. The licensee stated that for all systems that could transfer potentially radioa;tive liquids and gases out of the reimary 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.

Th'e 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 aut of service. When the safety features actuation signal clears, all valves

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. must be manual.l_y: repositioned.toctheir normal: position.- _FinallyL-t'he : licensee ]'

~ ta'ted that if-a containnient purge was-in progress, ::1e,: purge would be-termi-

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nated'by the-SFAS ror-by. a-hi'gh radiation signal-fre.rthe reactor building conitor..'Either_of these signals will secure the contlinment supply and

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. The subject of the containment isolation valves being cpened to allow purging during normal operation is rresently 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 staf" 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 s.aff.

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:

P verification, by inspection, of the operability of redundant a.

safety-related systems prior to the removal of 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, y

g-c.7;a meani i>f not'i?9ing ~ihvoTveirsadoCoper' ling pershiin51

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a whenever a rsafety-related system is-removed from and returned.

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.The. licensee's response to item 10 of this bulletin is documented in Reference 2 to'this evaluation.

With regard to subparagraph a above, the licensee stated that an Outage Coordinator, who maintains a Senior Reactor Operator's License, reviews any work and properly indicates the logging of tests for redundant safety-related equipment as P00RORGEL

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. required in the Technical Specifications (TS).

The TS for Rancho Seco require the testing of the redundant component (s) prior to maintenance on a cc.mponent(s) in a safety-rciated system.

In addition, the TS list the safety-related systems for whica 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 critaria are met.

In response to the Commission's 'Irder of May 7,1979, the licen.see has revised its procedures to require two operators to perform and verify, by signature, correct valve alignment of tne 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.

i'he 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 ens.ur.es.that_all. on-duty operators know when systerrs are in:or'out.of service.

. __.. _ 1

~ ~

.. ~ _ _ _.

The NRC staff finds ~ the }icensee's responses to item k of IE Bulletin 79-03E-

~

complete..

Ithbll:

" ll' opera'ing arid maintenance personnel should be made aware of the extreme seriousness and consequences of the simultaneous blocking of both auxiliary feedwater trains at the Three Mile Island Unit 2 plant and other actions taken during the early phases of the accident."

?00RBR8NM

. ~...

- 15 The licensee has documented in Reference 2 to this evaluation that all operating personnrel have had training on this matter and are aware of the extreme serious-ness and consequences of the TMI-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 would be acceptable, provided that the instructi.ons given to maintenance perscanel stressed the reasons why they do not have this authority and the potential consequences of incorrectly assuming this author.ity.

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 or everexposure, fire, sabotage or plant evacuation the Rancho Seco Emergency Plan Implementing Procedures require NRC notification. -

'-- A ra - ?.. m-

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- + _. +.

- The NRC staff finds the licensee's response to item 12 of IE Bulletin 79-05A complete; however, IE Bulletin 79-058 exoands the. licensee's responsibility in

~ t}iis area.-

Further discussion of this matter can ~ be. found under the staff's

~

evaluation of iten 6 of IE Bulletin 79-053.

. EVAL 0ATION dF RESPONSES TO IE BULLETIN 79-058'...... _.

. Ital:

" Develop procedures and train operation personnel on methods of establishing and maintaining natural circulation.

The procedures and training must include means of monitoring heat removal efficiency

_300ROR8jN/1 216-7 02c

. by available plant instrumentation.

The procedures must also contain a method of ass'uring 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 respense.

The instructions should include the following precautions:

A.

maintain pressurizer level sufficient to prevent loss of level indication in the pressurizcr; b.

assure availability of auequate capacity of pressurizer heaters, for pressure control and maintain primary system pressure to satisfy the subccoling criterion for natural circulation; and maintain pressure / temperature envelope within Appendix G limits c.

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. Coo 1down") was being revised to include methods of estab-

~

lishing and maintaining na~tural cir~culation including the monitoring of' heat

~

removal efficiency and subcooling criteria described in the bulletin.

However, the licensee stated' thi(adt'i_cipatoryif_il.1,ing 'of. the steam generittor was not expected to be necessary but that B&W was in the process of performing a review of-this area.

In addition, the licensee stated the procedure had been reviserto reflect the jirop'er~operato'r response upon loss of both main and

.... - -. ~. _

. auxiliary feedwater f1 s while in the natural circulation moda.

In Reference 5, the licensee reporte: 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 S.

The Office of Inspection and Enforcement (IE) reported in Reference 12 that the abcVe 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 79-053 ccmplete.

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:

-Operators do not override automatic action of engineered

.a.

safety features,, unless continued coeration ef enoineered a

safety features will reiult in unsafe olant conditions.

^

~

Fop eximole. if centinued coerati'on of encineersd safety

~

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. fea$ures wouTd threaten Eesctor vessel 'intecrity th'e HPI' ~

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Operating procedures currently, or are _.____.=:- - ~~

2._=. 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:

2 10.22..

. (1) Both low pressure injection (LPI) pumps are in operation and flowing at a rate in excess of 1000 gpa 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 subcoolina beyond 50 decrees F and the lencth of time HPI is in coera-tion shall be limited by the cressure/temoerature considerations for vessel intecrity. '"

• rNOTE:

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, subsequent to the issuance of IE Bulletin 79-058, 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 sp1cific emergency procedures for the Rancho Seco facility. The present HPI termination criteria,'as' defined in Re,ision 13 to Emergency Procedure 0.5 (" Loss'of Reactor Coo 1~ ant / Reactor-Cool, ant S.ystem Pressure) are as follows:.

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- :-. _ r "With HPI"autom tically actua.ted dua to low.RCS. pressure, 00,.N0T terminate a

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' 1). Both 1ow press. ur,adnfaction_(LP.I),_pumi,s are _ in operation _ and flowing

.at a-rate:in excass_of 4000 gpm each and the situation has been stable for 20 minutes; 2L67 023 P00R BRBlWL e%mm m

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All hot and cold leg temperatures are at least 50 below the saturation temperature for the existing RCS pressure and the hot leg temperatures are not more than 50* hotter thr.n the secondary side saturation temperature.

If 50 subcooling cannot be maintained, the HPI shall be reactivated.

RCS relief valve operations will prevent RCS pressure from exceeding 2750 psig.

(F.CS pressure recorder will indicate pressure instability s 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 likelibcod o~f automatic actuation of the pressurizer PORV during anticipated transients. This anclysis 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 P0RV for the spectrum of anticipated transients discussed by B&W in Enclosure 1.

Changtes developed by this analysis shall not result in increased frequency of pressurizer safety valve operation for these anticipated transients."

The licensee documented-i.ts response _~to this.. iter:1 off the.. bulletin in References 4 and 5 to this evaluation.

The licensee stated that it had reviewed the design aspec M of the rea'ctor protection system and-operating. procedures. which could, affect automatic actuation of the pressurizer pilot operated relief valve. _

~

The review included the analysis of anticipated transients performed by B&W.

The 1.icensee's. analys i.s _ concluded.that lowening_ the..high_ pressure. reactor _. trip._ _ ___.

.setpoint from 2355 psig_ta 2300 psig,..and. raising the pilot. operated relief-valve setpoint from 2255 psig to 2450 psig provided the requested reduction in automatic PORV actuation during anticipated transients.

Changes were incorporated

?00R ORJM 2167 o24

. into the appropriate plant operati 1rocedures.

In addition, the licensee stated that the setpoint changes 1_

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 ccmpleted by IE and reported in Reference 12.

The NRC staff finds the licensee's response to item 3 of IE Bulletin 79-053 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.

loss of offsite power; e.

low OTSG level; and, f.

l'os pressurizer level."

The licensee has documented its response to this item in Reference 5 to this evaluation. The licensee stated that procedures were ecdified to require prompt manual trip of the reactor upon:

loss of main feedwat er, turbine trip, loss of offsite power, icw steam generator water level and low pressurizer level.

Since the Rancho Seco design does not include main steam isolation valves,litsm"4.c of this bulletin is n.ot considered applicable.

~

L-In reviewing the requiremen.ts of the bulletin, it is.noted that this item lists 1ow pressurizer;leye_1_as an_ eiamp1.e_ of "transie_nt_s that resu1_t in a, pressure increase in the leagor. cool,a,nt syste,m." However, a low pressurizer level may-result from low reactor _ coolant _ system. pressure, and a prompt manual reactor trip-in some. casas _ may not necessarily be the mo.st advantageous, action for the operator to perform (i.e., An ov2rcooling event will cause both pres-surizer level and RCS pressure to decrease.

In this case, a reactor trip 2167 025 Ns] JR,Q&L

. 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 providir.g an autcmatic reactor trip for certain of these transients.

The NRC staff finds the licensee's response to item 4 of IE Bulletin 79-05B 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, turbirre 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.

Ihstallation of the modification would occur during the first

' refueling outage-followi.ng. receipt crf the required equipment.

~

{

. Subsequent to the. issuanca of IE Bulletin 79..05B,,,the Comm.ission. Ordar ;of.

. May_7., _.197E vas issued to the dicensee ;. 0ne..of the immediate actions required

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~

of the licensee, based on this Order, was to install a control grade reacter trip _for loss of main fqq.dwat_er_ and, turbine. trip._,The Order.requi?es that for ' _~}

d continued long-term operation, the licensee must: upgrade this circuitry to meet safef.y grade criteria.

A letter was issued to f.he licensee, dated September 7,1979 (Reference 14), which forwarded a request for additional P00R ORGNAL

. g 29

. 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 its 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 Balletin 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.

Item 6:

"The actions required in item 12 of IE Bulletin 79-05A are modified 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 r,esponse to this iten is contained in Reference 5 to this.

evaluation. This response was further clarified in Reference 16.

In its response, the licensee committed to infoming 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 Ope ations 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 conditions 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

~

an unplanned cessation oEreadtor cablant frow'eccursi~

~-

In addi-tion, to the above-cequirementsJ, any5ime the reactor is in an 'off noi-sl transi ent con-dition, the Shift Supervisor will notiff ~one of four designated management of ficials: --(l) Manager ofIuYiear '0[erftions~, (2) Plant Superintendirif,?('3T &

~

' Engineering and Quality Control Supervisor or (4) Operations Supervisor. These designated individuals hold SenioF ReacEcr Operators Licenses. Based ugon the management official's evaluation of the situation, he will _determ ne NRC report-i ability and take the necessary actions, one of which will be having the reporting Shift Supervisor use the dedicated telephone to notify the NRC of the situation.

. P00RORGNAL 2m e

. The licensee has installed a speaker phone in the control room, which is connected to the NRC Operations Center. The speaker phone will allow the control roca oper-ators to keep the NRC informed of any actions taking pla.e to cope with a plant emergency. The open communications c:iannel 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 79-053 compl e te.

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 cculd possibly be changed as a result of implementing the requirements of IE Bulletin 79-05B. 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 Bullatin would be appropriate when the plant modifications were made.

9y Reference 12 to this evaluation, which ' forwarded the staff's evaluation of the 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 reactorttrip. - These-TS changes were forwarded to the NRC for review by Reference'T to"this ' evaluation.

These prop ~osied TS changes are

.presently under review and wil'1 tie'.the subject oft separate safety ~ evaluation-

= --

.which is baing developed to s'upport the nedesiiry--Iidifise amenYme~nT.~ " ~

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.w The NRC staff finds the'N. _,'c e n s'de "s' res;ionse to 'iteT7 of IE Bulletirr>79-053

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complete.

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P00RBRGINAL e

• SU.' WARY OF OUTSTANDING ITEMS As a result of the staff's review of the licansee's responses to IE Bulletins79-05A and 79-053, the staff has identified gertain items for which additional information must be obtained in order to resolve these matters.

• /he list below summarizes these matters.

A more detailed discussion of these items is provided under the appropriate IE Bulletin item 7f this evaluation.

IE BULLETIN 79-05A Item 1:

The staff believes that a te.nperature monitor alone, downstream of the PORV is not always a valid indication of PORV position, and that a more direct means of monitoring the PORY position should be avail-able.

This item will be resolved as part of the licensee's compliance with Sec' tion 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 regi!;res 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:

Item 4.c required certain actions be taken by the licensee with respect to operation of the reactor coolant pumps following HPI initiation.

The requirements of IE Bulletin 79-05C supersede the requirements of item 4.c of this bulletin.

Responses to IE Bull.etin 79-05C are presently under review by the staff.

Following

. compi.etion of this re'/iew, the_.sta_ff_ will publish _a_. separate. evalua-tion _ covering the matter for all PWRs. _This evaluatica is-scheduled to-be published la~ter this year a.s.a NUREG documen.t.

Therefora,. na

- - - ' additional action]ill be taken on~this matter under IE Bulletin 79-05A.

Item 6:- 'Upon automatic initiation 'of safety injection, all lines whose

~~

isolation does~not' degrade cor6 cooling capability are. isolated for

~

the Rancho Seco containment with the exception of the lines necessary to assura continued operation of the RCPs and control rods. The

?00R OGM acceptability of not isolating these lines will be reviewed as part of the NRC staff's evaluation of the licensee's cc::pliance with Section 2.1.4 of NUREG-0578 (Section Title

" Diverse and More Selective Containment Isolation Provisiens for PWRs and BWRs").

Therefore, no additional action will be taken on this catter 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 lic.ensee's implementation of IE Bulletins79-05A and 79-05B and the Commission Order of May 7,1979. Therefore, no additional action will be taken an 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 aiid Ef'ects 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 under IE Bulletin 79-05A.

IE BULLETIN 79-058

,- 7,.

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Item 5:

This item requf_r_ed theq,icensee to. submit a proposed design a,nd.;

schedule for,a safety, grade, aptomatic anticipatory. reactor trip...

The Commission Order of May 7,1979, requires that this feature be installed as part of the long-term i'equirements of the Order.

Therefore, any additional informati_on-required of.t..%1.icensee in this matter will be reviewad as part of the NRC staff's evaluation of the licensee's compliance with long-term portion of the Order and r.o additional action will be taken under IE Bulletin 79-058.

~

2167~030

?DDR OR GINAl.

27-Item 7:

A separate evaluation to support the recuired license amendment will ba prepared that will include the licensee's proposed modifications to the Technical Specifications.

Therefore, no additiona' action will be taken on this matter under IE Bulletin 79-05B.

CONCLUSIONS Based on our review of the information provided by the licensee in response to IE Bulletins79-05A and 79-058, 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 THI-2 accident as they relate to the Rancho Seco, Nuclear Generating Station.

Th.ese actions have resulted in added assurance for the continued protection of the public health and safety during plant operation.

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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. Enge_lken (NRC), dated April 16, 1979, providing responses to itet~ 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 clarific' tion 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 'E 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 79-053 and providing documentation that item 3 of IE Bulletin 79-053 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 79-053 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 to items 5 and 7 of IE Bulletin 79-053.

9.

Letter from J. J. Mattimoe (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, forwarding Reportable Occurrence - R0-78-01 for Rancho 'Seco.

' l1.

Letter-from J. - J. Mattimoe (SMUD) to R_. H. Engelken '(NRC), Edited January 26,

~. -

1979, forwarding Report.able 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 wit.N the terms of.

the Order of May 7,1979 and. enclosing the " Evaluation of. Licensee's Compliance with the.NRC Order Dated May 7,1979 - Sacramento Municipa]

Utility District - Rancho Seca Nuclear Generating Station

' Docket No. 50-312."

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- 13.

NUREG-0578, "TMI-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 -com J. H. Taylor (B&W) to D. F. Ross (NRC), dated August 17, 1979, /crwarding B&W's generic report BAW-1554 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 79-053.

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