ML20002C063

Responds to NUREG-0737.Provides Info Re Shift Technical Advisor,Guidance for Evaluation & Development of Procedures for Transients & Accidents,Procedures for Feedback of Operating Experience & Training for Mitigating Core Damage

ML20002C063
Person / Time
Site:	Pilgrim
Issue date:	01/05/1981
From:	Morisi A BOSTON EDISON CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.1, TASK-1.C.1, TASK-1.C.5, TASK-2.B.4, TASK-TM 81-01, 81-1, NUDOCS 8101090045
Download: ML20002C063 (33)

Text

BOSTON EDISON COMPANY OsmatmAL Omete 500 SovLaTow Svatty Sc eTO N MAESACMUSETTS O219 9 A. V. M O RISI MANAGER NUCLEAR OPERATIONE SUPPORT DEPARTMENT January 5,1981 BECo. Ltr. *81-01 Mr. Darrell G. Eisenhut, Director Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555 License No. DPR-35 Docket No. 50-293 Resoonses to NUREG-0737 January 1, 1981 Reouirements

Dear Sir:

You have requested licensees of operating plants to submit documentation in accordance with schedule and criteria established in NUREG-0737. " Post T'il Requirements". Attached you will find our response to the following 0737 requirements.

I.A.1.1 Shift Technical Advisor i

I.C.I Guidance for the Evaluation and Development of Procedures for

Transients and Accidents I.C.5 Procedures for feedback of operating experience I.C.6 Verifying Correct Performance of operating activities II.B.2 Design Review of Plant Shielding and Environr.antal Equipment Qualification II.B.4 Training for Mitigating Core Damage j II.E.4.2 Containment Isolation Dependability II.F.2 Instrumentation for Detection of Inadequate Core Cooling.

II.K.3.3 Reporting Safety and Relief Valve Failures and Challenges l

II.K.3.13 Separation of HPCI and RCIC Initiation Setpoints ,g II.K.3.17 ECCS Outage Report l II.K.3.21 Restart of Core Spray & LPCI Injection Systems II.K.3.22 Auto Switch,ver of RCIC 81010900'6 [

TOETON EDIRON COMPANY I

i tir. Darrell G. Eisenhut, Director

January 5,1981 Page 2  !

II.K.3.44 Evaluation of Anticipated Transients with Single Failure to Verify No Fuel Failure II.K.3.45 Depressuri 3 tion other than full ADS III.D.3.4 Control racm Habitability

Several of our responses were developed in conjunction with the BWR Owners' Group and General Electric. These positions were not available for review until December 10, 1980, at which time an in-house review was commenced. As a result of our review, further development is recuired to satisfy plant specific requirements at Pilgrim Nuclear Power Station in response to II.K.3.13.

This work has begun and a detailed response will be provided.to your staff by January 31, 1981.

4 We trust this letter is resconsive to your requirements; however, should you desire additional information or clarification, please feel free to contact us.

I Very truly yours, i

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1.A.I.1 Shift Technical Advisor i

I In response to previous NRC requirements contained in NRC letter dated September 13, 1979 and October 30, 1979, Boston Edison Company has created the A description of the rosition of Shift Technical Advisor (STA) at PNPS Unit .#1.

! si program was provided in BECo letter #80-54, dated April 4,1980. A copy of eqr response is attached for your convenience. As we indicated in our Cecember 15, 1980 submittal as a result of a reorganization within the nuclear organiza-

tion, the STA group no longer reports to the Assistant Station Manager. Presently, STA's report to the Staf f Assistant-Nuclear Safety, who in term reports to the Nuclear Operations Manager, in response to your requirement for a description of our STA training, the followino is nrovided:

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The STA Training Program has been developed using as a guide the recommen-

! dations contained in the INPO document entitled " Nuclear Power Plant Shif t Tech-nical Advisor - Recommendations for Position Description, Qualifications, Educa-l l

tion and Training", Revision 0, dated April 30, 1980. The course content for the initial group of STA candidates is as follows:

I A. Seven weeks (approximately 260 contact hours) of training on-site, adminis-tered by the Pilgrim Nuclear Power Station Training Group. This training

! includes:

Plant Specific System Training covering all the BWR Specific Systems of Soc-

tion 6.4 of the INPO Occument.

Plant Specific Reactor Technology i

Plant Chemistry j

Nuclear instrumentation and Controls Reactor Plant Thermal Cycle Process Instrumentation and Control Review of Reactor Theory including Reactivity Control, Reactivity Coef fi-cients, and Fission Product Doisons Review of Radiation Protection and Health Physics S. Three weeks (120 contact hours) of simulator training, administered by a vendor on a SWR simulator. This training includes 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of time on the simulator and 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> of classroom time. Primary emphasis is on reactor operations, totn normal and abnormal, and response of the plant to transi-ents and accidents. .

C. Eight weeks (approximately 304 contact hours) of training, tailored specif!-

cally for Pilgrim STA's and administered by General Electric Company, on the ,

following subjects:

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. t Contact Hoars t

Station Nuclear Engineering 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> Abnormal Event Analysis 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> Nuclear and Non-Nuclear instrumentation 4C hours Control Room Management 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> Communications / Motivation 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> j j D. One week (40 contact hours) on Mitigation of Core Damage and Emergency Pro-cedures, administered by various BECo specialists. ,

! By January 1,1981, al l initial STA candidates will have completed the train-Ing outlined under A and B above. it is the position of Boston Edison Company that the training provided as of January 1,1931 and the qualifications of the j

STA candidates, each of whom holds a Bachelor's Degree or equivalent, are suffi-  !

cient to demonstrate conformance with the criteria of your October 30, 1979 letter, particularly in view of the fact that as of January 1, 1981, all STA candidates will have had at least 520 hours0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br /> on-duty experience serving the STA function. The j l

balance of the STA training is presently scheduled to be completed by June I, 1991.

! Plans for requalification training will be in place by January 1,1982.

1 Scheduling of STA training has been impeded by an Industry-wide crunch on hiring qualified STA candidates, conflicting regulatory requirements to both train i

STA's and have an STA on duty on each watch, and lack of availability of simulator i time and quallfled training instructors.

In response to your requirement for a description of our long term STA pro-

gram, the following is provided

The aim of the long term STA program is to provide an on-shif t technical advisor to the Nuclear Watch Engineer, who has college level education In engineering and science subjects and specific training in the response and analysis of the plant for transients and accidents, until such time as these characteristics are attained by the Nuclear Watch Engineer. The STA is accountable for the following end results:

A. Contributes to maximizing safety of operations by independently observing f

plant status and advising shift supervision of conditions that could compro-mise plant safety.

B. Contributes to maximizing plant safety during transient or accident situations by independently assessing plant conditions and by providing the technical assistance necessary to mitigate the incident and minimize the ef fect on per-sonnel, the environment, and plant equipment.

These accountabilities are consistant with those in the referenced INPO docu-ment. Boston Edison's long term STA program closely follows that-described in the (NPO document. Boston Edison agrees with the Commission's assessment that -

the descriptions as set forth in Sections 5 and 6 of Revision 0 to the INPO docu-ment are an-acceptable approach for the selection and training of personnel to staff the STA position. Hcwever, it is our intent to use this document as a guide-line only and to deviate from it at our discretion, so long as, in our Judgement,

' such deviation does not substantially af fect the quality or Intent of the STA pro-gram.

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-3 i For example, we would contemplate covlation, on a candidate by candidate basis, 4

in the following subsections of the INPO cocument:

5 5.2 Excerience - The document specifies at least 12 months experience shall be at tne stattsn at which the cosition is to be filled. All of our present STA candi-dates met.. this requirement. However, considering the dif ficulty we have experi-l' enced in tilling job vacancies and the potential high turnover rate for the STA cosition, we intend to maintain the option to waive this recuirement.

5.3 Absences f rem STA Duties - The document requires certain training prior ta f assuming on-shift responsibilities for any STA who has not been actively perform- i ing that function for a period of 30 days or longer. Boston Edison agrees that a the person absent f rom STA cuties for periods greater than 30 days shall be briefed on significant procedure and facility changes during that absence. Absences from l

STA s If ts as part of the STA training program or as part of the of f-shif t STA function, including the operating experience assessment function, will not be [

j spplicable to this definition of absence as these are considered to be integral to the STA function.

6.0 Educatien and Traininn Recuirements - Section 6.1 of the document specifies certain prerequisite education and callege level fundamental education considered necessary for successful completion of the advanced course work specified in Sec-tions 6.2 through 6.8. It has been our experience that capable, highly motivated S

Individuals can, in fact, pass the advanced course work without formal training in f

' certain prerecuisite or fundamental subjects, by initiating self-study programs or j tutoring, as necessary. The selection criterion for assuming the STA function is based on successful completion of oral and written exams given along with the advancec course work. It is Boston Edison's position that candidates who pass these comprehensive exams, including certification, if required, have demonstrated an

! acceptable level of knowledge of the prerequisite subjects and need not document i formal training in these areas. L However, in recognition that certain of the fundamentals are more important than others to the STA function, the extent of the advanced course work training for

- Boston Edison STA's has been increased to 19 weeks as compared to 15 weeks speci-l fled in the INPO document, with heavy emphasis on nuclear eng!neering, abnormal I event analysis and mitigation of core damage. Also, the cualifications for an STA candidate recuire the Individual to have a Bachelor's Degree in engineering or science plus 3 years power plant experience or equivalent. Or equivalent is defined as an Associate Degree in nuclear or nechanical engineering plus 5 years nuclear experience.

As regards licensing plans and plans for phase out of the STA program, these matters are still under review. The STA's serving in that position will be encouraged to attain SRO licenses, but inis will be on a voluntary basis. Phase ,

out of the STA program is dependent upon the success of the program to upgrade the educational background of current watch Engineers. Because the program for the Watch Engineers is developmental at this time, the actual duration of the STA pro-gram is not determinable. It is an objective of the STA program Standards to prepare STA's, of increased who are amenable, for future positions as Watch Engineers.

operational experience and NRC licensing will be part of the development of those l

STA's inclined to progress to that position. Those STA's who may obtain NRC licenses, but are not attracted to the Watch Engineer position may move to other positions in the nuclear organization when and if the'STA program is phased out.

As stated !n our December 15, 1980 submittal, we are reevaluating and will reissue our position related to Technical Specifications changes for this require-ment, l

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l Attach-ont to 1,A,.,,

2.2.1.b Shift Technical Advisor The Boston Edison Company has created the position of Shif t Technical Advisor (STA) at F11 grim Station Unit #1. This position is accountable for the safe operatich of Tilgria Unit il through contributing to assesscent of plant con-ditions during normal operation and transients consistent with technicalThe Shift Tech crecifications, procedures and regulatory requireeents. maintains direct line reporting Advisor Croup is independent of Operations andThe Shift Technical Advisor is assigned to to the Assistant Station Manager. One (1) STA is on a specific shift and there are three (3) shifts per Jay.The STA maintains a high duty at all times during nor=41 plant operation.cvareness of safety in pl plant conditions during operation The STAand transienta.

provides expertise the perspective and to the Watc l t safety.

diagnose and respond to unusual events.the t, we for assessing plant condition On a daily basis, the STA co==unicates with other AlsoSTA's to report on a daily basis,ongoing plant conditions and the status of any special circu= stances. h ffectiveness the incu= bent (s) interface with the Watch Engineer (s) to assess t e e of operations.

As required, the incu= bent (s) provide technical i f Operating coc=en plant Engineer.

operations and responses of operatora to transients for cvent to the On-Site Technical Support Center The scope Staff.is of the pos- essentially o face with the Watch Engineer regarding plant operations. sted ition also includes providing rece==endations to the Watch cvent or trancient.

Boston Edison believes that our program meets the intent of NUREG-0578 Po 2.2.1.b, Shift Technical Advisor.

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1.C.1 Guidance for the Evaluation and Development of Procedures for Transients and Accidents in a letter cated June 30, 1980 (letter #MFN-117-80, R. H. Buchholz to D. G. Eisenhut), G.E. submitted SWR Emergency Procedure Guidelines on behalf of the ShR Owners Group. Boston Edison is a participant in the eWR Owners Group and endorses this submittal. Following review of the guidelines and issuance of further guidance by the NRC, Boston Edison intends to implement these guidelines in emergency procedures at FNPS 1 on a schedule consistant with the requirements of item 1.C.I.

1.C.5 Procedures for Feedback of Ooeratino Experience to Plant Staff Procedures governing feedback of opera +1ng experience have been complated, and w1II be in effeet by January 31, 1981.

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1.C.6 Guidance en crocadores for Vert ' vine Correct Perfor-3nce of C?eratine Activities The re;uirements of TAP lte. 1.C.6 are teing reviewed against existing sta-tion eclicies and procedures. Where differences are identified, a decisicn will be made wnether er not to incor? crate the change. It is expected that this effcet aill te complets1 witn the necessary crecedure changes by June 1931.

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11.B.2 Desinn Review of Plant Shieldinn and Environmental Qualification of Ecole-ment for Scaces/ Systems which may be used in Post Accident Ocerations in a letter dated April 4,1980 (BECo letter #B0-54, G. C. Andognini to H. R. Denton) Boston Edison described the shielding design review performed at PNPS 1 and indicated potential system modifications and shielding add'*lons under evaluation to increase accessibiltry to plant vital areas following tne post-ulated accident. This reviE3 was performed to the requirements of NUREG-0578, item 2.1.6b.

On the basis of clarification provided in NUREG-0737, item 11.B.2, however, Boston Edison is reviewing this initial study and reevaluating previous results.

A preliminary reanalysis of the Control Room, Post Accident Sampling location and Sample Analysis area indicates each would provide the required level of accessi-biliTy following the postulated accident, it is anticipated that reanalyses of other plant vital areas will result in similar findings. Modificrtions in pro-gress or presently scheduled for completion by January 1,1982 which are related to insuring vital area accessibility following an accident are as follows:

1. Remote Closure Capability for Reactor Building Truck Lock Door as des-cribed in Boston Edison's April 4, 1980 letter (Item 2.1.6b).

2. Post-Accident Sample Sink Installation as required by NUREG-0737, item 11.B.3 and described in Boston Edison's April 4, 1980 letter (item 2.i.8a).

3. Remote operation capability for Post Accident Combustible Gas Control valves as described in Boston Edison's April 4, 1980 letter (Item 2.1.5a).

Regarding deviations f rom position 11.B.2, Boston Edison does not intend to consider a "LOCA event in which the primary system may not depressurize" in deter-mining dose rates. Emergency procedures in effect at PNPS 1 require that the RCS be promptly depressurized and cooled down with low pressure systems following an accident of large scale fuel damage.

11.B.4 Traininq for Miticatino Core Damace A program has been developed at PNPS 1 to teach the use of installed equip-ment and systems to control or mitigate accidents in which the core is severly damaged. This program is intended for training STA's and operating personnet f rom the plant manager through the operating chain to the licensed operstors and includes all training indicated in enclosure 3 to H. R. Cenion's March 28, 1980 letter. The initiation of this training program is tentatively scheduled for February 1931, with the initial program scheduled for completion by April 1, 1931.

11.E.4.2 Isolatien Docendabilitv in response to previous NRC requirements contained in NRC letters, dated September 13 and October 30, 1979, Boston Edisen has reviewed the containment isolation systen in accordance with criteria established by the NRC requirements.

Boston Edison believes its position contained in BECo letter #80-54 to be respon-sive to NRC positions 1-4 of Task Action Plan item 11.E.4.2. This position is attached for your convenience.

in response to position 5 of 11.E.4.2 to reduce the containment setpoint pressure that initiates containment isolation for nonessential penetrations to the minimum compatible with normal operating conditions, the following is provided:

The SWR Owners Group and General Electric have prepared a response to this NRC requirement which demonstrate the adequacy of the present containment isolation serpoint of approximately 2 psi. Boston Edison endorses this owner's group pos-ition. In addition, BEco letter #80-57 requested and provided justification to raise the High Drywell Pressure trip level setting from 2 psig to 2.5 psig. NRC letter dated May 12, 1930 approved this change to PNPS's Technical Specifications.

Bosten Edisen believes its present Hign Crywell Pressure 2.5 psig is adequate and no further change is required. ,

in response to position 6 of 11.E.4.2, which requires containment purge valves that do not satisfy the operability criteria set forth in Branch Technical Position CS136-4 or the Staff Interim position of October 23, 1979 to be sealed closed as defined in SRP6.2.4, the following is provided:

Boston Edison in response to the staf f interim position implemented controls to satisfy this requirement, which were reviewed and approved by your staf f in a letter dated Sectember 9, 1930.

Subsequently, we have not been in compliance to limit the operation of the containment vent and purge valves to less than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> a year. We are presently develocing f urther procedural controls to prevent reoccurence of this condition which will be complete by January 15, 1981. We believe, this will satisfy the staf f's interin position for containment purge and vent valve operation l

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, Page 1 of 9 Attachment to 11,E,.4,7 4,

• 2.1.4 Containeent Isolat ion Provisions for PVR's and EVR's_

Fosition

1. All containment isolation system designs shall co= ply with the recoceendations of SRP 6.2.4; i.e. that there be diversity in the para =eters sensed for the

. in.i.tiation of contain=ent isolation. ~

2. All plants shall give careful reconsideration to the definition of essential i

l and non-essential syste=s, shall identify each system determined to be esstntial, shall identify each system determined to be non-essential, shall describe the basis for selection of each essential syste=, shall sodify their contain=ent isolation designs accordingly, and shall report the results of the re-evaluation to the NRC.

3. All non-essential systems shall be autocatically isolated by the contain=ent isolation signal.

4. The design of control syste=s for auto =atic contain=ent isolation valves shall be such that resetting the isolation signal vill not result in auto-tatic reopening of contain=ent isolation valves. Reopening of contain=ent isolation valves shall require deliberate operator action.

Response

1. Diversity of para =eters sensed for the initiation of containuent isolation shall be provided in accordance with SRF-6.2.4.

Isolation Diversity in Para =eters Sensed for Initiation of Contain=ent A. Secondary Contain=ent Isolation (TSAR 5.3.3.3)

Either of two signals vill initiate the secondary contain=ent system.

These signals, which indicate a loss-of-coolant accident inside In the dryvell are high ciryvell pressure or low reactor water level.

addition, radiatian monitors in the operating (refueling) floor ventil-stion exhaust duct, which indicate a fuel handling accident, can initiate the secondary contain=ent system. Secondary contain=ent can also be initiated manually from the control room.

B. Table 1 su==arizes the isolation signal codes (asterisk ite=s only)Addit- used by the Primary Containment and Reactor Vessel Isolation System.Section 7 3.4.7 (Isolation ional details may be found in PNPS 1 FSAR Tunctions and Settings)

Exceptions to the diverse isolation signals criteria The NRC have hasbeen accepted identified the to the NRC in response to IE Bulletin 79-08.

existing methods for isolation of all valves except the reactor water sa=ple valves, the MSIV drains, and the RWCU supply and return valves.

• The reactor water sa=ple valves presently receive only one isolation sig-nal (Iow-low reactorisolation A second water level) signal that meetshigh containing the dryvell diverse isolacion pressure criteria.

vill be added to the existing logics to provide the diverse signals re-quired as these valves have no effect on plant safety.

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Changing the isolation signals to the HSIV drains, however, could effect plant safety. Operation of these valves to more restrictive isolating

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requirements than the HSIV's could possibly result in condensate buf! dup .

between the HSIV's thus preventing operation Either (opening) of the MSIV's orfailure i

dacaging the steac: lines to the condenser.clicinate the condenser as a

.one possible method of cool devn.

RWCU suctien and return Ifue isolation valves are currently,provided isoletten with

, signals.

only one contain=ent isolation signal in addition to the processT i the vessel water during the situation where high dryvell pressure ex sts because the dryuell coolers are not operating ble to or a s=all without reaching a low reactor vessel level condition. It is desira keep the RWCU operating under these conditions.

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Response

2. Definition of Essential and Non-Essential Svste=s A. Source of Definition is NL* REG 0578, Pg. A-14

3. Definitions:

These syste=s that should be selectively

1. Essential Syste=s:

isolated during contain=ent isolations only after it is estab-lished that the use of these syste=s vill not be =eeded for an accident or abnormal transients.

These syste=s not needed for mitigation

2. Non-Essential Syste=s:

i of tn accident or abnor=al transient and which should be i=ediately isolated during contain=ent isolation.

Criteria for Imple=enting Definitions:

C. FSAR (Sect. 1.5.2.6.2)

1. A prieary contain=ent shall be provided to co=pletely e

~ resetor vessel. following accidents that release It shall be radio-possible erial barrier during or active material into the pri=ary contain=ent. ih st at to test the primary contain=ent integrity and leak t g tne periodic intervals. in-I 2.

A secondary contain=ent that co:plet91y encloses both primary con ment and fuel storage ' areas shall b provided and s a act as a radioactive material barrier. h

3. The pri=.ary and secondary contain=ects, in conjunction f radioattive with ot er

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• engineered safeguards, shall act to prevent the release ideline o

• zaterial fro = the contain=ent volu=es free exceeding the gu values of applicable regulations.

i h Provisions shall be =ade for the re= oval of energy i of thefro = with n t e

4. to the pri=ary contain=er.t as necessary to maintain the l primary contain=ent.

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5. Piping that both penetrates the primary containment structure and l could serve as a path for the uncontrolled release of radioactive zaterial to the environs shall be automatically isolated whenever Such such uncontrolled radioactive material release is threatened.

isolation shall be effected in time to prevent radiological effects I

from exceeding the guideline values of applicable regulations.

- Classification of Systees l -

- Essential Systees_

a. RER (except head spray)

b. Standby Liquid Control

c. RCIC

d. Core Spray (except test lines)

e. HFCI

f. Mrin Stea: Flow Instru=entation 3 Urywell Pressure Instrumentation i

h. RECCW - see note

1. Contain=ent Atmospheric Control System Non-Essential Systems

a. NMin Steam

b. Feedwater

c. Reactor Water Sample

d. Control Rod Drive Hydraulic Return

e. Control Rod Drive Inlet and Outlet

f. RER Reactor Head Spray 3

Reactor Water Cleanup l

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h. Core Spray Test Line to Suppression Poel l

I 1. Drywell Equipment Drain l

j. Drywell Floor Drain i

l k. Traversing In-core Probe

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l 1. Service Air

m. Instrument Air Note: RSCCW has 2 Class C containment isolation valves (check valves

• motor operated gate valve), one valve per containment h pen

• zent but do not communicate directly with the reactor Class vessel..with C Ifnes t e primary contain=ent free space, or with the environs.

require only one valve which closes automatically by p w e In d & XIL

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Respense ,

3. All Non-Essential Systems Shall be Autoratica!!y Isolated by the Containeent Iselation Signal Table 1 gives a listing of all nor.-essential syste=s and their respective isolation signals. Items which require greater detail are described in

' the'next few paragraphs

_Tip Valves Sectien 5.2.3.5.2 of FSAR I

Tip system guide tubes are provided with an isolation valve which closes autocatically upon receipt of proper signal and after the TIP cable and fission chacber have been retracted. In series with this isolation valve, an additional or backup isolation shear valve is included. Both valves are located outside the dryvell. The function of the shear valve is to assure l

integrity of the contain=ent in the unlikely event that the other isolation valve should fail to close or the cha=ber drive cable should fail to retract if it should be extended in the guide tube during the tice that contain=ent isolation is required. This valve is designed to shear. the cable and seal the guide tube upon an actuation af gnal. Valve position (full open or full closed) of the automatic closing volves vill be indicated in the control room.

Each shear valve vill be operated independently. The valve is an explosive type valve and each actuating circuit is monitored. In the event of a con-tain=ent isolation signal, the TIP syste: receives a com=and to retract the traveling probes. Upon full retraction, the isolation valves are then closed automatically. If a traveling probe were ja==ed in the tube run such that it could not be retracted, instru=ents would supply this information to the operator, who would in turn investigate to determine if the shear valve should be operated.

Section 7.5.9.2.2. of PNPS 1. FSAR A valve system is provided with a valve on each guide tube entering the primary contain=ent. These valves are closed except when the TIP subsyste= is in operation. A ball valve and a cable shearing valve are mounted in the guide tubing just outside of the primary contain=ent. They prevent the loss of reactor coolant in the event a guide tube ruptures inside the reactor vessel.

A valve is also pro;ided for a gas purge line to the indexing mechanisms.

A guide tube ball valve opens only when the TIP is being inserted. The ahear valve is used only if a leak occura when the TIP is beyond the ball valve and The shear valve, which is controlled by a manually Power to the TIPS fails.

operated protected switch, can cut the cable and close off the guide tube.

The shear valves are actuated by detonation squibs. The continuity of the squib circuits is monitored by front panel indicator lights in the control room.

in the closed position.

• A guide tube ball valve is normally de-energized and 9khen the TIP starts forward the valve is energized and opens. As it opens it actuates a set of contracts which gives a signal light indication at the TIPS control panel and bypasses an inhibit limit switch which automatically stops TIP motion if the ball valve does not open on com=and.

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Coerressed Air System Valves sectton 10.11.3.1 of PhTS 1 FSAR e

Fressure loss in the high pressure system, sensed by several pressure sv.sthes, vill cause valves in the service air header, the low pressure service air

. cross-around line, and the non-essential instrument air header to close in a casesding sequence thus leaving the essentisi instrument air header as the only header drawing air from the receivers in the event that supply pressure decreases. ,

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Page 6 of 9 Rerponse

4. Reopening of Containment Isolation Valves _*'

11, 1979 and March 18,1980, the During meetings with the NRC on DecemberNRQ accepted To the PNPS with the exception of the primary containnent vent and purge system.

reset the Bop isolation logic af ter a scram, it is first necessary to h 30preset the Cencral Electric isolation logic at panel C905 and then reset t eTh logics at panel C7.is the replacement of the existing reset pushbuttons on panel C selector switches, and this has been done.

I The control circuits for the primary containment vent and purge system iso-lation visves have been revised by viring valve control switch contacts l

(either directly or through auxiliary relays) parallel to the nor= ally clo f reset selector switch contacts.a control switch is in an open ii position,) untilprovide a pa trip relays energized for isolation (independent of reset switch pos t onAt that tine.

all control switches are moved to close.

r be reset by operation of the keylocked reset selector switch.

Also affected by NURIG 0578 were the MSIV's, the reactorofwater

" position the sa=ple valves the dryvell su=p effluent valves, and the "ezergency open vent and N2 makeup valves.

h The MSIV's control circuits were revised by viring "close" contacts This from eac "close" MSIV svitch in series with the applicable trip logic reset c

' before the trip logics can be reset after an automatic isolation.

2 The control circuits for the " emergency open" position licable of the vent and N control makeup valves have been revised by viring contacts from the app t avitches (via auxiliary relays) in series across the trip relay sealin contac . l This arrangement, similar to that used on the MSIV's, In addition, requires we that all va ve control switches be closed before the trip relay is reset. h NRC, to allow are replacing the existing control switchea as recom= ended i bykey t eto operation between the "Open" and "Close" without a key and requ re a get into the " Emergency Open" position. been The control circuits for the re=aining valves with reset h three position, have problems modified by replacing zaintained contact control sviches ll require witThese switches spring return to normal switches.

been reset.

auxiliary relays provide sealin circuits which,when tr

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.. Page 7 of 9 TABLE I s'50L ATl0N SIGNALS TO WJN-ESSENTI AL SYSTD#$

Isole* lon Stone ~l_

NarreI Sietus(1) i

$s sf e- Peaitret Ion Open 8,C,D,P,Q~ s la Main Steam Lines Closed (2)

B,C,D,P,Q b Main Stes. Drains Open Rev. Flow (Check Valves) 2 Res: tor Feed = ster Closed (2) B,C,D,P,Q,{A 3 Res: tor Water Sa.ple N:tc 4 Rev. Flow (Check Valves) 4 C C Return Note 4 Note 4 5 CC in and Outlet 6 RA Head Spray Closed %F Open A, W, Y, J, RM 7 Re:: tor hater Clasnup Closed (2) G B C5 Test Line 8, F Open 9 Drywell Equip. Drains 8, F Open 10 Drywell Floor Drains Closed (2)(3) FA lia TIP Pele.ary RM (Explosive Shear Valve)

Open b Backup Closed Rev. Flow (Che:k Valve) 12 53rvice Air Open inside - Rev. Flow (Che:k Valve) 13 Instru .ent Air Open Outside - N4 MTES: tion (1)

Norral status position of a valve is the position during norral power opera i

of Valvethe reactor.

can be opened or closed by renote manual switch for operating conven e (2) dur.ing any mode of rescior operation except when suice-4 tic signal is "A" or "F" causes autcr.atic withdra al of TIP probe, then valve (3) Sigpal (4) autor.stically closes by mechanical action.CC solenold valves are norr and during scram.

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' ' Page 8 of 9 ISOLATION SICNLL CODES FOR TABLE I i DESCRIPTION S IG'?4 L

• Ao Reactor vessel low vetor level - scram and close Isolation valves except r.aln steam lines.

Ikonctor vessel low low water level - Initiate R0lC, HP01 and close main l B' steam line Isolation valves.

C* High radiation - main steam line (also causes scram).

D8 Line break - r.aln steam line (steam line high space feeperature or high

steam flow).

l Reactor low low level or high drywell pressure - select LPCI and close t

l E other loop valves.

To High drywell pressure - close Rm/ shutdown cooling and head s; ray plus the RHR to rad asis valves.

l G Reactor vessel low water level and low pressure; or high dryvell pressure -

initiate Core Spray and RHR systems.

Jo Line break in cleanup system - high space temperature, or high flow.

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Line break In ROIC system steam line to turbine (high steam line space temperature or high steam flow) or low steam pressure.

LO Line break in HPCI system steam line to turbine (high steam line space temperature or high steam flow) or low steam line pressure.

Line break in RHR shutdown and head cooling (hlgh space temperature; M8 alarm only; no auto closure).

main steam line pressure at Inist to main turbine (R'JN mode only).

P' Low

) Lov drywell pressure - close containment spray valves.

S T

Low reactor pressure permissive to open core spray and RHR.LPCI valves.

U High reactor vessel pressure - close RHR shutdown cooling valves and l head cooling valves.

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• High temperature at outlet of cleanup system nonregenerative heat .

• exchanger. .

Y Standby liquid control system actuated.

Rmote manual swlich from control room. * '

RP.8 D"*0 @

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fage 9 of 9 L .

g Roactor high water level Isolete main steam line (except In run ecde).

'X RC'IC or HPCI steam supply valve (as applicable) not fully closed.

0These are the isolation functions of the primary containment and reactor vessel isolation control system; other functions are given for Inforr.ation only.

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11.F.2 Instrumentation for Detoction of Inadequato Core Coolinq Cn December 23, 1979, General Electric submitted on behalf of the BWR Owners Group (letter hvFN-314-79, R. H. Suchhol: to D. F. Ross) a prepublication version of Section 3.5.2.3 of NED0-24708 in response to NUREG-057S, item 2.1.3b. This submittal indicated that "the reactor vessel water level neasurement technique provided on the General Electric SWR performs satisf actorily for all modes of nor-mal operation, anticipated transient conditions and credible accident conditions".

In a later submittal (SECo letter #80-263, dated October 29, 1930) Boston Ediscn indicated its concurrence with the BWR Cwners Grcup position that existing instru-nentation is adequate for detecting inadequate core cooling.

As stated in the December 15, '980 submittal to the NRC, Boston Edison is in the process of reevaluating its position on this item on the basis of new informa-tion and clarification contained in NUREG-0737, item 11.F.2. The NRC will be not!-

fled prorotly of any changes to Eosten Edison's present position that installed instrumentation is satisfactory.

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11.K.3.3 Recortinn Safety and Relief Valve Failures and Challences l

Boston Edison has reviewed planr operating records since April 1, 1980 to identify all relief valve challenges or failures. We have defined a relief valve challenge to be anytime a relief or safety valve received a signal to operate via i High reactor pressure, auto signal (ADS) or control switch (manual). Based on 1 this definition, all relief and safety valve challenges at PNPS rasulted from

} manual actuation except for two occurances when a relief valve opened due to high nitrogen pressure to the relief valve solenold. The following is a list of relief

and safety valve challenges since April 1, 1980 (1980 refuel outage ended May 16, 1980):

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Challenge Date S/RV # Reason for Challenge Remarks 5/17/80 RV203-3A Performed tech spec sur- All valves tested OK 3B veillance l'

3C l 3D l

! 5/25/30 RV203-3D Perform oper test per temp. Valve failed to open l Procedure TP 80-65 Test after maintenance Valve would not open 5/25/80 RV203-3D i

1 Test af ter maintenance Valve would not open 5/25/80 RV203-3D i

i 5/25/30 RV203-3C Test after maintenance Tested OK l related to problems with RV203-3D 5/25/80 RV203-3B Test af ter maintenance Tested OK related to problems with RV203-3D 5/26/30 RV203-3A Test after maintenance Tested Ok related to problems with RV203-3D t

Test after maintenance Would not open 5/26/80 RV203-3D RV203-3D Test after maintenance Tested OK 5/26/80 r.

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Safety / Relief Valve Challenge S/RV d Peason for Challence Remarks Challenge Date RV203-30 Re-test (reliability) Tested OK 5/26/80 RV203-3A Operability Test Tested OK i S/1/80 RV203-38 Operability Test Tested Ok S/1/S0 RV203-3C Operability Test Tested OK S/1/30 Operability Test Valve did not open 8/1/80 Rv203-3C RV203-3D Operability Test after Tested OK S/3/60 maintenance 8/5/S0 RV203-3D Operability test 3 times Test OK 8 times RV203-3C Accelerated test program Tested OK S/30/60 Rx scran/ operator open & Valve opened-would not 10/1/80 RV203-3D closed valve via control close switch RV203-30 Test after maintenance Tested OK 10/5/30 10/7/80 Rv203-3A Valve opened due to hi instrument nitrogen pressure to the RV solenoid RV203-3A Test after :mintenance Tested OK 10/3/80 10/31/80 RV203-3A Valve opened due hl instrument nitrogen pressure to the RV solenoid .

11.K.3.13 Separatien of WPCI and RCIC System Initiation levels In conjunction with the SAR Owners Group and General Electric, TAP ltem 11.<.3.13 was addressed as tso separate issues wnich are: a) Separation of HPCI and RCIC initiation levels and c) Auto restart of the PClC system on low water level, in response to the separation of HPCI and PCIC Initiatien level, General Electric en behalf of the BWR Owners Group submitted a position def ending the present initiation levels of the HPCl and PCIC system. This was transmitted to the NRC in G.E. letter, dated October 1, 1990. Boston Edison endorsed this pos-ition in BECo letter #S0-246, dated October 1,1930. Boston Edisen believes the response provided by the Owners Group and endorsed by us adequately addresses the NOC concern and no further action is required.

General Electric on behalf of the SWR Owners Group evaluated the second NOC re;uirement to modify RCIC logic to have it restart en subsecuent low water level signals. G.E. evaluation showed that this change would contribute to incroved system reliability and that it could be acccmplished without adverse effect en system function and plant safety. Bosten Edi son has reviewed G.E's proposed mod-ification and generally concurs with G.E.'s solution for the addition of auto reset to the RCIC system, heaever, a detailed proposal based on plant s:ecific recuirements of PNP 3-1 must be solicited f rom G.E. and examined by EECo before a commitment is made to incorporated any mcdificatien to the RCIC system. We antici-pate to have a description of the proposed modification no later than January 31, 1981, l

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11.K.3.17 Report on Outaces of ECC Svstems and Procosed Technical S0ecification Chances SECo contracted General Physics Corporation to prepare the required ECCS outage report. Work commenced in October and Involved a significant amount of record searching, technical specification review, and review of post oper-ation logs. The draft report submitted to SECo was confusing and hard to Inter-pret, therefore, after further revision, the ECCS outage report will be submitted by January 15, 1931.

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I 11.K.3.21 Restart of Core Soray and LPCI Systems The SWR Owners Group and General Electric have reviewed this NRC requirer.ent and co not believe the NRC suggestions will necessarily enhance the safety of the plant. This conclusion is based on the adequacy of the current ECCS logic design coupled with the potentially negative impact on overall safety of the proposed changes.

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11.K.3.22 Automatic Switchover of RCIC System Suction--Verify Procedures and Modify Cosien noston Edison has reviewed station operating procedures to verify that pro-cedures exist clearly describing manual switchover c." the RCIC suction f rom the Condensate Storage Tank to the Suppression Pool. The necessary procedures will be revised by January 15, 1981.

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11.F.3.44 Evaluation of Anticloated Transients with Sincie Failure to Verifv no Fuei Faifure The SWR Owners Group and General Electric have reviewed this NRC requireNnt.

It was shown that for SWR /2-6 plants 1 adequate core cooling is ra.aintained for the

' norse case conditions evaluated. Boston Edison has reviewed the Owners Group pos-tilon and believes It to be applicable to PNPS Unit 1 No further action is required.

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11.K.3.45 Evaluation of Cecressurization with other than ADS The BWR Owners Group and General Electric have prepared a report in resconse to TAP lten 11.K.3.45. The report shows that depressurization rates other than full ADS:

1) Do not exceed vessel integrity limits for a ful f ADS bicwdown.

2) For slower depressurization rates, there is little imcact on vessel fati-gue relative to full ADS biowdown.

3) Slower depressurization rates have an adverse impact on core cooling cap-ability.

Boston Edison has reviewed the cwners group report and endorses the position.

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111.C.3.4 Control Room Habitability Pecoirements Bosten Edison has ccmpleted a control rocm habitability study ir acccrdance with criteria established in NRC TAP ltem 111.D.3.4 The results of the study demonstrate that PNPS centrol room operators are adequately protected against the effects of accidental release of toxic and radioactive gases and that the Nuclear power plant can be safely operated or shutdown under design basis accident condi-tions. (GDC19). This study will be forwarded to your staff under se arate C0ver by January 21, 1991.

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