ML18082A280

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Viewgraphs Entitled, NRC Staff Review of Facility for Fuel Load & 5% Power Ol
ML18082A280
Person / Time
Site: Salem PSEG icon.png
Issue date: 04/16/1980
From:
NRC COMMISSION (OCM)
To:
Shared Package
ML18082A278 List:
References
NUDOCS 8005010162
Download: ML18082A280 (42)
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Text

NRC STAFF REVIEW OF PSE & G's *

  • SALEM UNIT 2 FOR FUEL LOAD
  • AND 5o/o POWER OPERATING LICENSE

CHIUIL1:5Tt>N M~.$S.

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T ISOHI j

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L ACTION PLAN ITEMS REVIEWED BY JOINT NRR/IE TEAM Shift Manning Organization and Management Criteria Safety Engineering Group Other Items Resident Inspector Licensee Dissemination of Operating Experience

  • Applicant/NRC Communications STATUS Review Team Visited PSE & G March 5-7 Final Staff Position is Presented in SER Supplement No. 4

ORGANIZATION AND MANAGEMENT*

Findings

1. Integrated organizational arrangement under the Senior Vice President -

Energy*

Supply and Engineering.

2. Clear lines of management control from the Senior VP to the operating unit and the technical support organizations.
3. Effective lines of communication between the operating unit and the technical
  • support elements..
4. Management involvement in and acceptance of responsibility for matters affecting plant safety.
5. Provisions for effective management oversight of nuclear operations.
6. Well qualified individuals in the management chain.

Conclusion Corporate management is sufficiently qualified and involved to assure a continual understanding of plant conditions and safety considerations.

.. -. *-~*-----***-

ONSITE SAFETY ENGINEERING GROUP Requirement Establish an independent group, assigned onsite, to perform independent evaluations of the technical adequacy of all important safety-related procedures and changes and evaluation and assessment of the plant's operating experience and performance.

Finding The applicant will provide an onsite group of five engineers to perform this function.

This group will perform independent reviews of station operating activities and will, include the following disciplines -

nuclear engineering, heat transfer, mechanical engineering and instrumentation and controls. This group will report off-site.

Conclusion The applicant's proposal is acceptable.,

SHIFT MANNING AND EXPERIENCE NRC POLICY PSE&G PSE&G PROPOSAL For two unit operation with separate control rooms

  • 2 Senior Reactor Operators (-SRO)

(must be licensed on both units)

  • 4 Reactor Operators (RO) (2 assigned to each unit and licensed on that unit)

Has agreed to comply with these requirements, however, they presently have only the following operators licensed:

Number 8

4 20 11 Unit 1 Operating 1 SRO*

2 RO Type of License Unit 1 and 2 SRO Unit 1 and 2 RO Unit 1 SRO Unit 1 RO Unit 2 Fuel Loading 1 SRO* (Shift supervisor licensed on both units) 1 SRO* (in charge of fuel loading) 1 RO** (in control room)

  • SRO's to work 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />shifts.
    • Will limit Unit 2 to modes 5 and 6 until additional RO's are licensed on Unit 2.

PSE&G LICENSED OPERATOR EXPERIENCE SRO's -

Average experience of licensed operators -

3 years Minimum experience of licensed operator -

1.5 years RO's Average experience of licensed operator -

2.5 years Minimum experience of licensed operator -

1.0 year

f.

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QUALIFICATION OF REACTOR OPERATORS SECY 79-330 E APPLICABLE TO SALEM UNIT 2 COLD APPLICANTS REQUIREMENT IMPLEMENTATION SALEM

1. PREVIOUS EXPERIENCE FOR ASAP MEET REQUIREMENTS SENIOR OPERATORS
2. SHIFT TRAINING EXPEDITIOUS MEET REQUIREMENTS
3. INSTRUCTOR QUALIFICATIONS EXPEDITIOUS ALL INSTRUCTORS HAVE, OR HAD, SENIOR LICENSE
4. SPECIFIC EXERCISES ON EXPEDITIOUS MEET REQUIREMENTS
  • SIMULATOR
5. INCREASE SCOPE OF ASAP MEET REQUIREMENTS EXAMINATIONS PARTIAL

. 6. SIMULATOR EXAMINATION ASAP DO NOT MEET REQUIREMENTS

7. PASSING GRADE INCREASED ASAP MEET REQUIREMENTS
8. INFORM MANAGEMENT OF ASAP MEET REQUIREMENTS EXAMINATION RESULTS

.... *--* ----** --------------- ---~~~.___..

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QUALIFICATIONS OF R.EACTOR OPERATORS SECY 79-330 E NOT APPLICABLE TO SALEM COLD APPLICANTS REQUIREMENTS IMPLEMENTATION SALEM

1. HOLD OPERATOR'S LICENSE FOR
  • ASAP MEET REQUIREMENTS ONE YEAR PRIOR TO SENIOR LICENSE
2. SIMULATOR TRAINING FOR
  • EXPEDITIOUSLY MEET REQUIREMENTS HOT APPLICANTS
3. NRC AUDIT TRAINING PLANNED
4. RETRAINING ON SIMULATOR ASAP PLANNED
5. NRC REOUALIFICATION PLANNED EXAMINATIONS
6. REVISE ANS 3.5 EXPEDITIOUSLY IN PROCESS (SIMULATOR STANDARD)
7. NRC EXAMINER SELECTION AND ASAP IN PROCESS TRAINING STUDY
  • ~**:....
            • ,~. ;..,~.. -...,.... *............ * *** **......

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SHIFT TECHNICAL ADVISOR (STA)

Graduate Engineer EDUCATION:

. TRAINING:

35 week program during 1980

  • Reactor Theory and Thermodynamics
  • Reactor Operations
  • Health Physics and Chemistry
  • Reactor Systems
  • Accident Analysis
  • Reactor Simulator
  • Metallurgy PREVIOUS POWER PLANT EXPERIENCE:

CURRENTLY ASSIGNED ST As:

PREVIOUS NUCLEAR EXPERIENCE:

CURRENTLY ASSIGNED ST As:

33 months to none At least one year 21 months to none At least one year PERFORMS ACCIDENT ASS.ESSMENT:

Yes PERFORMS OPERATIONS ASSESSMENT:

. Yes REPORTS TO:

Station Reactor Engineer REOUALIFICATION:

On an annual basis OTHER:

May serve a dual role as a test engineer during low power testing.

I.

i I

I I

EMERGENCY PLANNING REQUIREMENTS FOR LOW POWER TESTING

  • 10 CFR PART 50, APPENDIX E
  • REGULATORY GUIDE 1.101
  • NUREG 75/111 & SUPPLEMENT 1

........,. -........... _.. ~.........

EMERGENCY PLANNING STAFF FINDINGS

MEET REG. GUIDE 1.101 (SER SUPPLEMENT 3, 12/78)

DELAWARE CONCURRENCE (6/78)

FEMA LETTER (3/80)

  • PLANS ACCEPTABLE FOR FUEL LOADING AND LOW POWER TESTING
  • FULL POWER REQUIREMENTS INTERIM CRITERIA (NUREG 0654)

NRC LETTER (3/80)

STAFF EVALUATION OF SALEM 2 CONTROL ROOM BACKGROUND

  • PSE&G REVIEWED UNIT 2 CONTROL ROOM -

IDENTIFIED SOME DEFICIENCIES

  • NRC/ESSEX CONDUCTED DETAILED REVIEW/AUDIT MARCH 19-25 STAFF FROM l&C CONDUCTED DETAILED INSPECTION OF CONTROL ROOM PANELS INTERVIEWS WITH OPERATORS WALKTHROUGHS OF EMERGENCY PROCEDURES CONCLUSIONS
  • CONTROL ROOM ACCEPTABLE FOR LOW POWER TESTING
  • DEFICIENCIES IDENTIFIED WHICH REQUIRE CORRECTION PRIOR TO EXCEEDING 5% POWER

DEFICIENCIES IDENTIFIED DURING NRC/ESSEX CONTROL ROOM REVIEW

  • lAMP TEST FEATURES
  • PROCEDURES
  • VERTICAL METER FAILURE
  • LABELING I

CORRECTIVE ACTIONS REQUIRED

  • INCREASE AUDIBLE ALARM LEVEL
  • INSTALL LAMP TEST RECEPTACLE
  • RETYPE PROCEDURES (LARGER TYPE) INCLUDE REFERENCED MATERIAL WITHIN EOP
  • OPERATOR AWARENESS OF MID RANGE FAILURE
  • PERMANENTLY ATTACH COMPONENT LABELS

UNDERCLAD CRACKING IN REACTOR VESSEL NOZZLES I. Underclad Cracking History II. Underclad Cracking Mechanisms

. Ill. Cracking Experience for Near Term Ols IV. Evaluation of Cracking for Near Term Ols

I. UNDERCLAD CRACl<ING HISTORY A. Reheat Cracking (1970)

  • Found in Nozzle Forgings Clad in Europe
  • Affected Fabricators Rotterdam, B & W, GE, CBI, CE, Sulzer
  • Cracking Potential in Nozzles Clad Before 1973
  • Maximum Crack Size: 1 /8" Deep x 1 /2" Long B. Cold Cracking (1979)
  • Found in Nozzle Forgings and S/G Tube Sheets Clad in Europe
  • Affected Plants Prairie Island 1 and 2, Sequoyah 1, McGuire 2, Catawba 1, Watts Bar 1 and 2, Offshore Power
  • Maximum Crack Size:*

Vessel Nozzle, 1/4" Deep x 3/4" Long SIG Tube Sheet, 1/2" Deep x 1-1/4" Long,

HEAD CLOSURE FLANGE VESSEL WALL THICKHESS-TRAHSITION THICKNESS= 8.625" IN-CORE INSTRUMENTATION PENETRATIONS

. VESSEL WALL TO LOWER HEAD THICK-NESS TRAMS IT ION OUTLET NOZZLE

CRACKS CLA.D DEPOSITION I INTER FA.CE CIRCUt\\t\\fERENTIAL DIRECTION l

e e-e 8 *=

8 e

8

~

AXIAL DlRECTION CLAD REHEAT CRACKS 8

COLD CRACKS

  • 11. CAUSES OF UNDERCLAD CRACKING
  • Cladding Process Pre Heat -

Clad Deposition -

Post Heat A. Reheat Cracking

  • Single Clad Layer -

High Heat Input

  • Cracking in Local Degraded Zone at Interface of Adjacent Clad
  • Depositions
  • . Remedial Measures Multi-Clad Layers Lower Heat Input B. Cold Cracking *
  • Low Pre and Post Clad Deposition Temperature
  • Remedial Measures Maintain Adequate Pre and Post Clad Temperatures

Ill. CRACKING EXPERIENCE IN NEAR TERM Ols A. Sequoyah 1

  • Potential for Reheat and Cold Cracks
  • Some Cracks Detected in Each Nozzle
  • Reheat and Possible Cold Cracks B. Watts Bar 2
  • Potential for Cold Cracks
  • No Cracks Detected C. North Anna 2
  • Potential for Reheat Cracks
  • Many Cracks Detected in Each Nozzle
  • Reheat Cracks D.* Salem
  • Reheat and Cold Cracks Not Expected
  • Some Cracks Detected in 4 of 6 Nozzles
  • Unidentified Cracking Mechanism

r

-~--.--

IV. EVALUATION OF CRACKING FOR NEAR TERM Ols

  • No Exact Correlation Between Crack Presence and Crack Mechanisms
  • Cracks Sizes Within ASME Code Section XI Acceptance Standards
  • Cracks are Acceptable for Service

IE BULLETIN 79-27 11LOSS OF NON-CLASS IE l&C POWER SYSTEM BUS DURING OPERATION"

  • Concern: Oconee Unit 3 Event -

November 10, 1979

  • Action Required:
  • Evaluate Effects of Loss of Power to l&C
  • Propose Design Modifications
  • Change Emergency Procedures
  • Licensee to Respond by July 1, 1980
  • Preliminary Staff Evaluation Indicates:
  • W Design Relies on Protection System for Control
  • Redundant Protection System Measurements Displayed in Control Room
  • Single Power Failure Won't Cause Loss of All Status Indication Required for Recovery
  • Resolve Issue Prior to Escalation Above 5 % Power

ENVIRONMENTAL QUALIFICATION

  • Barton Lot 1 Transmitters (Pressure, Level, Flow).
  • Qualified for FSAR Accident Environment
  • Replace Units After Second Fuel Cycle*
  • Main Coolant Temperature Sensors
  • Qualified for 10° Rads
  • Acceptable for One Fuel Cycle (One Year Normal and Accident Dose)
  • Replace or Requalify for Higher Dose
  • "Interim Staff Position on Environmental Qualification of Electrical Equipment", NUREG-0588
  • Applicant to Address Deviations and Justify
  • Staff Review of Exceptions and Basis *
  • Requalification if Additional Margin of Safety Required
  • Test Results Indicate Acceptable Short-Term Operation (Trip Function); Tests Did Not Include Additional Margins to Prove Long-Term Post-Accident Monitoring Capability.

FIRE.PROTECTION PROGRAM SALEM UNITS 1 & 2

  • FIRE PROTECTION EVALUATION COMPLETED
  • UNITS 1 & 2 ARE ESSENTIALLY IDENTICAL AND HAVE BEEN REVIEWED TOGETHER
  • BOTH UNITS WILL MEET APPENDIX A TO BTP-APCSB 9.5-1
    • THERE ARE NO OPEN ITEMS
  • 32 MODIFICATIONS HAVE BEEN COMPLETED
  • AN ALTERNATE SHUTDOWN SYSTEM TO ACHIEVE HOT SHUTDOWN IS INSTALLED
  • REMAINING 5 MODIFICATIONS, INCLUDING AN ALTERNATE SHUTDOWN SYSTEM TO ACCOMPJJSH COLD SHUTDOWN WILL BE COMPLETED BY OCTOBER, 1980

r UNITED STATES

  • NUCLEAR REGULATORY COMMISSION.
  • fa/***

WASHINGTON, D. C. 20555 April 9, 1980 MEMORANDUM FOR:

Chairman Ahearne

  • Commissioner Gilinsky Commissioner Kennedy Commissioner Hendrie Commissioner Bradford FROM:

Harold R. Denton, Director:

THRU:

SUBJECT:

Office of Nuclear Reactor Regulation William J. Dircks,

//fJ~ _£_ 0. S::P Acting Executive Director for Operations CONSIDERATION OF OPERATING LICENSE FOR SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 The staff has completed its review of Salem Nuclear Generating Station, Unit No. 2 for issuance of a 5% power license.

O~r review is doc~mented ih the enclosed draft Supplemental Safety Evaluation Report (SSER).

It includes disposition of the residual pre-TMI issues (some cf which were discussed in our SER cf October 11, 1974 and Supplements l, 2 and 3 thereto) and the list cf requirements for Near-Term Operating Licenses (NTOL) which was

  • based on consideration of the actions contained in the NRC TMI Acti*on Plan.

Regarding the Salem Unit 2 control room, the NRC staff, together with its consultants from the Essex Corporation, conducted a 5-day preliminary on-site review of the control room on March 19-25. The NRR *Instrumentation and Control Systems Branch participated in the review.

The objective of this revi~w was to assass the degree to which the current design reflects specific human factors considerations. The review included the control and display design and location; workstation layout, including visibility and reach envelopes; control room environmerit, specifically noise and illumination; and design and content of emergency procedures.

r~

In general the control room was designed to promote effective and efficient

~°"

operator actions.

The controls and displays are functionally grouped and

~I generally well integrated. In addition, the Unit 2 control room is isolated

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from the Unit l control room by two glass part;tions and a central corridor *. J- ~

This isolation aids in reducing noise and control room traffic.

f"'

f\\.1 J' ~

We did not identify any control room items requiring completion prior t~ ~ ~

issuance of a 5% power license.

Our evaluation of the control room is. 'f-

~v ffi:~

included in the enclosed supplement.

l~ \\

Concerning the shift manning needed to man Units No. l and 2 for all modes o;'-' ~

operation, the applicant does not presently have adequate licensed operators to permit Unit 2 operation to 5% power with Unit 1 operating.

We therefore will limit Unit 2 operation modes 5 and 6 pending the licensing of additional Unit 2 operators.

Our evaluation of this matter is also contained in the enclosed supplem~nt.

.r

  • The Commission

---.-~*-:--

  • *~:::.:-_.*_*

...... On October 18, 1979, a Section 2.206 petition was filed by Alfred c. and Eleanor G. Coleman requesting among other things a stay of the licensing of Salem Unit 2.

The *petition raised the concern that the staff had not analyzed the effect of the operation of Salem Units 1 and 2 on the shortnose sturgeon, which is listed by the Secretary of Interior as an endangered species. In compliance with this agency's obligations under Section 7 of the Endangered Species Act, the

{NMFS) on October 31, 1979.

On December 7, 1979 the NMFS issued a biological opinion with respect to the effect of operation of Salem Unit 1 on the short-nose sturgeon.

The staff was requested by NMFS to provide more information with respect to the effects of operation of Salem Unit 2 and construction of

  • Hope Creek on the shortnose sturgeon *. Information i ndi eating that the future operation of Salem Unit 2 will not jeopardize the continued existence of the shortnose sturgeon was provided to NrfS in the form of a biological assessment.

on March 14, 1980.

The staff is presently awaiting the issuance of a biological opinion by NMFS with respect to the operation of Salem Unit 2 and the construct-ion and operation of Hope Creek Units 1 and 2.

NHFS personnel have orally in*

dicated that they concur with the staff 1 s conclusion.

However, the staff does not believe that the Salem Unit 2 license should be issued before official receipt of this biological opinion.

~e are prepared to meet with the Commission on the operating license for Salem 2 at your convenience.

cc w/encl~: SECY OGC OPE

Enclosures:

Harold R. Denton, o*;rector Office of Nuclear Reactor Regulation

1. Draft Supplement No. 4 to the SER
2.

Proposed License

'I,*'.)

WASHlNGTC.N. D. C. 20~*55 PUBLIC SERVICE ELECTRIC AND GAS COi*~PANY PHILADELPHIA ELECTRIC COMP~NY DELMARVA PO\\>JER AND LIGHT COl-iPANY ATLANTIC CITY ELECTRIC crnwANY SALEM NUCLEAR GENERATING STATION, UNIT 2 DOCKET NO. 50-311 LICENSE FOR FUEL LOADING AND LOW POWER TESTING License No. DPR-75

1.

The Nuclear Regulatory Commission (the Commission) having found that:

A.

The application for license filed by the Public Service Electric and Gas Company for itself and the Philadelphia Electric Company, Delmarva Power and Light Company and Atlantic City Electric Company (herein after referred to as the licensees) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's regulations set forth in 10 CFR Chapter I and all required notifications to other agencies or bodies have been duly made; B.

Construction of the Salem Nuclear Generating Station, Unit 2 (facility) has been substantially completed in conformity with Construction Permit No. CPPR-53 and the application, as amended, the provisions of the Act an9 the regulations of the Commission; C.

The facility requires an exemption from certain requirements of Appendix J to 10 CFR Part 50.

~his exemption is described in the Office of Nuclear Reactor Regulation's Safety Evaluation Report, Supplement No. 4.

This exemption is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest.

The exemption is, therefore, hereby granted.

With the granting of this exemption, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.

.* D.

The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Cwrnission; E.

There is reasonable assurance: (i) that the activities authorized by this license can be conducted without endangering the health* and safety of the public, and (ii) that such activities will be conducted in compliance with the regulations of the CoCTnission; F.

Public Service Electric and Gas Company is technically qualified to engage in the activities authorized by this license in accordance with the regulations of the Commission; G.

The licensees are financially qualified to engage in the activities authorized by this license in accordance with the regulations of the Commission;.

H.

The licensees have satisified the applicable prov1s1ons of 10 CFR Part 140, 11 Fi nanci al Protection Requirements and Indemnity Agreements", of the Commission's regulations; *.

I.

The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public; J.

After weighing the en vi ronmenta l, economic, technical, and other benefits of the facility against environmental and other costs and considering available alternatives, the issuance of License No. DPR-75 subject to a condition for protection of the environment set forth herein is in accordance with 10 CFR Part 51, of the Cormnission's regulations and all applicable requirements have been satisfied; and K.

The receipt, possession, and use of source, byproduct and special nuclear material as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40, and 70, including 10 CFR Sections 30.33, 40.32, 70.23 and 70.31.

2.

Pursuant to approval by the Nuclear Regulatory Comoission at a meeting on April

, 1980, License No. OPR-75 is hereby issued to the Public Service Electric and Gas Company, Philadelphia Electric Company, Delmarva Pm1er and Light Company, and Atlantic City Electric Company (licensees) to read as follows:

A.

This license applies to the Salem Nuclear Generating Station, Unit 2, a pressurized water nuclear reactor and associated equipment (the facility), owned by the licensees. The facility is located in Salem County, New Jersey, and is described in the 11Final Safety Analysis Report" as supplemented and amended (Alllendments 10 through -44) and the Environmental Report as supplemented and amended (Supplements l through 3).

B.

Subject to the conditions and requirements incorporated herein, the Commission hereby licenses:

(1)

Public Service Electric and Gas Company, pursuant to Section l04b of the Act and 10 CFR Part 50, to possess, use, and operate the facility at the designated *1ocation in Salem County, New Jersey in accordance with the procedures and limitations set forth in this license; (2)

Public Service Electric and Gas Company, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3)

Public Service Electric and Gas Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumen-tation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

Public Service Electric and Gas Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, *source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5)

Public Service Electric and Gas Company, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the f aci 1 ity.

C.

This license shall be deemed to contain and is subject to the conditipns specified in the following Commission regulations in 10 CFR Chapter I:

Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or h2reafter in effect; and is subject to the additional conditions specified or incorporated below:

(l) Maximum Power Level Public Service Electric and Gas Company is authorized to (a) load fuel, (b) proceed to initial criticality, (c) perform startup testing at zero power in Operational Mode 2, and (d) after prior written approval by the Director of Nuclear Reactor Regulation, operate the facility for testing at reactor core power levels not in excess of 171 megawatts thermal (five percent of rated power).

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B attached hereto have hereby been incorporated into this license.

Public Service Electric and Gas Company shall operate the faci li.ty in accordance with the Technical Specifications.

( 3)

Secondary \\,!at er Chemistry The licensees shall implement a secondary water chemistry monitoring program to inhibit steam generator tube degradation.

This program shall include:

a.

Identification of a sampling schedule for the critical parameters and control points for these parameters;

b.

Identification of the procedures used to measure the values of the critical parameters;

c.

Identification of process sampling points;

d.

Procedure for the recording and management of data;

e.

Procedures defining corrective actions for off control point chemistry conditions;

f.

A procedure identifying (i) the authority responsible for the data and (ii) th~ sequence and timing of administrative events required to initiate corrective action; and

g.

Monitor the condehsate at the effluent on the condensate pump.

When leakage is confirmed repair or plug the leak in accordance with Branch Technical Position MTEB 5-3.

(4)

Initial Test Program

( 5)

The licensees shall conduct the post-fuel-loading initial test program (set forth in Section 14 of the Final Safety,ll,nalysis Report and its Amendments) that has been revi e1'.'ed and approved by the Commission at the time of issuance of this license l'lithout making any major modification of this program. Major modifications are deerned to involve unreviewed safety questions under 10 CFR Section 50.59 and are defined as:

a.

Elimination of any test identified in Section 14 of the Final Safety Analysis Report as essential.

b.

Modification of test objectives, methods or acceptance criteria for any test identified in Section 14 of the Final Safety Analysis Report as essential.

c.

Performance of any test at a power level different from there described.

d.

Failure to complete any tests included in the described program*

(planned or scheduled for power levels up to the authorized power level).

ui red to The following conditions shall be completed to the satisfaction of the Commission prior to fuel loading.

The following conditions are related to matters specified in the TMI Action Plan, Near Term Operating License (NTOL) Requirements, dated February 6, 1980, and applicable to fuel load and zero power testing. Each of the follow-ing conditions references the appropriate section of Part II of Supplement No. 4 to the Safety Evaluation Report (NUREG-0517 Suppl.

No. 4) for the Salem Nuclear Generating Station, Unit 2 and follows the nurrbering sequence utilized in the February 6, 1980 NTOL Requirements list.

6 -

a.

Shift Technical Advisor (I.A.1.1)

Public Service Electric and Gas Company (PSE&G) shall provide on each shift a Shift Technical Advisor (STA) whose principal duty shall be to act as an advisor to the shift supervisor primarily in assessing accident/tran-sient occurrences.

The Shift Technical Advisor (STA) shall be a degreed engineer possessing specialized training as specified below.

During 1980, interim Shift Technical Advisors (STA) shall serve on shift. During 1980, these interim STAs shall receive training in plant systems including mechanical and control systems, thermal hydraulic, core*design, technical specifications and transient and accident analyses.

Also during the same period of 1980, designated permanent STAs shall undergo an approximately 35 week training pro-gram.

Training shall be provided in reactor theory and thermodynamics, reactor operation, health physics and chemistry, reactor systems, accident analysis, reactor simu-1 a tor and metallurgy.

PSE&G shall develop shift turnover procedures which assure transferral of i nfomat ion between STAs.

b.

Shift Su ervisor Duties I.A.1.2) and Shift Personnel Res onsibilities I.C.3 PSE&G shall issue a management directive, signed by the General Manager, Electric Production, which emphasizes the assignment of primary management responsibility to the senior shift supervisor.

Thi~ directive shall be reissued on an annual basis.

The responsibility of the shift supervisor shall be that defined in PSE&G Administrative Procedure No. 5, 110perating Practices".

The senior shift supervisor or shift supervisor shall remain in the control room area at all times when Unit 2 is operating in Model, 2, 3, or 4, except that he may be allowed to be absent provided an individual, other than the shift technical advisor, who possesses a valid senior reactor operator (SRO) license assurr.es the control room co:11rnand function during his absence.

The individual who assu~es the control room command function shall remain in the control room until the senior shift supervisor or shift supervisor returns and reassum2s the com~and function.

The senior shift supervisor or shift supervisor shall only be relieved by an individual who possesses a valid SRO license.

Individuals who do not possess a valid SRO license, including members of station management, shall not relieve the. shift supervisor, nor shall they direct the licensed activities of licensed operators.

c.

Shift Manning (I.A.1.3)

The PSE&G's shift crew composition for the operation of Salem Unit 2, in conjunction with Unit 1 shall be in accordance with Table 6.2-1 of the Technical Specifications.

PSE&G shall have administrative procedures that include provisions which limit the amount of overtime worked by licensed operators.

The need for a licensed operator to exceed the limits in overtime shall be infrequent. The limits on overtime hours are:

(1)

An individual should not be permitted to work more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> straight (not including shift turnover time).

(2)

There should be at least a 12-hour break between all work periods (shift turnover time is included in this 12-hour break).

(3)

An individual should not work more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7-day period.

(4)

An individual should not work more than 14 consecutive days with out having two consecutive days off.

- 8..

However, for those circumstances which arise requiring deviation from the above, such deviation may be authorized by the plant manager or high levels of PSE&G management in accord-ance with est ab 1 i shed procedures and with appropriate documenta-tion of the cause.

d.

Revised Scope and Criteria for Licensing Examinations (I.A.3.1)

PSE&G shall train all shift personnel (licensed and non-licensed) at Salem Nuclear Generating Station, Units l and 2 in all the TMI related design changes which have been incorporated in the

~ station.

e.

Organization and Management Criteria (I~B.1.1)

With respect to offsite technical support to the plant staff in the event of an emergency, PSE&G shall develop procedures delineating the responsibility and authority of corporate office personnel in providing technical support.

f.

Safety Engineering Group and Onsite Evaluation Capability (l.B.1.2)

PSE&G shall establish an onsite Safety Engineering Group which shall be technically responsible to offsite manage-ment.

The Safety Erigineering Group shall be independent of the station staff and shall perform independent reviews of station operational activities. This group shall be composed of five persons; one supervisor and four additional pers ans with collective expertise in the areas of nu cl ear engineering, heat transfer, mechanical engineering and instrumentation and controls.

The duties and responsibilities of this group shall be as follows:

1.

Be cognizant of the scope and intent of the test program.

2.

Be cognizant of the tests being conducted.

3.

Be familiar with the operation of a pressurized water type reactor.

4.

Provide daily status reports to the Manager of ~uclear Operations at the corporate office.

PSE&G shall establish an onsite capability to evaluate operating history of Salem Unit 2, in conjunction with that of Salem Unit 1, and also pl~nts of similar design.

g.

Lic-ensee Onsite Operating Experience Evaluat1on (I.B.1.4 and License Dissemination of 0 I. c. 5 PSE&G shall develop procedures for assessing and disseminating operating experience to o~erators and other personnel involved with plant operation, ~oth in plant and at PSE&G's corporate office.

h.

Shift Relief and Turnover Procedure (I.C.2)

PSE&G shall implement the shift and relief turnover procedure in accordance with PSE&G Administrative Procedure AP-5 "Operating Practices", Operating Memo 20, and in the Operations Department Manual.

PSE&G sha 11 est ab 1 i sh a system to eva 1 ua.te the effectiveness of the turnover procedures.

i.

Control Room Access (I.C.4)

PSE&G shall implement control room access procedures in accordance with Administrative Procedure AP-5, "Operating Practices 11 Lines of comrnuni cation and authority for p 1 ant management personnel not in direct command of operations, including those who report to stations outside of the control room.

shall be in accordance with the Emergency Plan Manual.

j.

Degraded Core Training (II.B.4) z PSE&G shall initiate preparation of a program to ensure that all operating personnel are trained in the use of installed plant systems to control or mitigate an accident in which the core is severely damaged.

The training program shall include the following topics:

Incore Instrumentation

1.

Use of fixed or movable incore detectors to determine extent of core damage and geometry changes.

2.

Use of thermocouples in determining peak temperatures; methods for extended range readings; methods for direct readings at terminal junctions.

Excore Nuclear Instrumentation (NIS)

1.

Use of NIS for determination of void formation; void location basis for NIS response as a function of core temperatures and density changes.

Vital Instrumentation

1.

Instrumentation response in an accident environment; failure sequence (time to failure, n1ethod of failure);

indication reliability (actual vs. indicated level).

2.

Alternative methods for measuring flows, pressures, levels, and temperatures.

a.

Determination of pressurizer level if all level transmitters fail.

b.

Determination of let down fl ow with a c 1 ogged filter (low flow).

c.

Determination of other Reactor Coolant System parameters if the primary method of measurement has failed.

Primary Chemistry

l.

Expected chemistry results with severe core damage; consequences of transferring small quantities of liquid outside containment; importance of using leak tight systems.

2.

Expected* isotopic breakdo\\'m fo.r core damage; for cl ad damage.

3.

Corrosion effects of extended irrrnersion in primary water; ti me to fa i1 u re.

  • Radiation Monitoring
1.

Response of Process and Area Monitors to severe damages; behavior of detectors when saturated; method for detecting radiation readings by direct measurement at detector output (overranged detector); expected accuracy of detectors at different locations; use of detectors to determine extent of core damage.

2.

Methods of determining dose rate inside containment from measurements taken outside containment.

Gas Generation

1.

Methods of H2 generation during an accident; other sources of gas (Xe, Kr); techniques for venting or disposal of noncondensibles.

2.

H2 flammability and explosive limit; sources of Oz in containment of Reactor Coolant System.

k.

Relief and Safety Valve Test (II.D.2)

PSE&G shall carry out a testing program to qualify the relief and safety valves under expected operating conditions for design basis transients and accidents in accordance with the requirements and schedule provided in NUREG-0578, Section 2.1.2, as clarified in NRC letter to operating license applicants, dated November 9, 1979.

1.

Relief and Safety Valve Position (II.D.5)

PSE&G shall provide reactor system relief and safety valves with a positive indication in the control room derived from a reliable valve position detection device or a reliable indication of flow in the discharge pipe. The indication shall comply with the requirements contained in NUREG-0578, Section 2.1.3.a, as clarified in NRC letter to operating license applicants, dated November 9, 1979.

m.
  • Auxiliary Feedwater Initiation and Indication (II.E.1.2)

PSE&G shall provide for automatic initiation of the auxiliary feedwater system and shall provide for indication, in the control room, of auxiliary feedwater flow to each steam generator. These requirements shall comply with NUREG-0578~

Sections 2.1.7.a and 2. 1.7.b, as clarified in NRC letter to operating license applicants, dated November 9, 1979.

n.

Inadequate Core Cboling - Subcooling ~eter (II.F.2)

PSE&G shall provide a subcooling meter to provide on-line indication of coolant saturation condition. The meter shall complj with the requirements of NUREG-0578, Section 2.1.3.b, as clairfied in NRC letter to operating license applicants, dated November 9, 1979.

o.

Inadequate Core Cooling - Additional Instrumentation (II.F.2)

  • PSE&G shall provide a design of additional instruments to provide an unambiguous indication of inadequate core cooling.

This requirement shall comply with NUREG-0578, Section 2. 1.3.b, as clarified in NRC letter to operating license applicants, dated November 9, 1979.

p.

Emergency Power for Pressurizer Equipment (II.G)

PSE&G shall provide emergency power for the pov1er-operated relief valves (PORVs), the PORV block valves and pressurizer level instrument channels. This requirement shall comply with NUREG-0578, Section 2.1~1, sub-section 3.2, as clarified in NRC letter to operating license applicants, dated November 9, 1979.

q.

IE Bulletins on Measures to Miti ate Small Break LOCAs and Loss of Feedwater Accidents II.K.

PSE&G shall review th~ operating procedures and operator training related to the measures to mitigate small-break LOCAs and loss-of-feedwater transients. These requirements shall comply with Items l through 13 of IE Bulletin No.79-06A, as modified by IE Bulletin No.79-06C.

The requirements of Item 7(b) of IE Bulletin No.79-06A have been superseded by the requirements set out in Section 2.0.6.a of this license.

r.

Improve Licensee Facilities for Responding to Emergencies -

Onsite Technical Support*Center (III.A.l.2(a))

s.

PSE&G shall establish an onsite technical support center to meet the January 1, 1980 requirem2nts of NUREG-0578, Section 2.2.2.b, as clarified by NRC letter to operating license applicants, dated November 9, 1979.

PSE&G shall provide an onsite operational support center to meet the requirements of NUREG-0578, Section 2.2.2.c, as clarified by NRC letter to operating license applicants, dated November 9, 1979.

t.

Emeraenc Pre aredness of State and Local Governments -

Near-Term Actions III.Band III.B.l During the period of this license, PSE&G shall maintain in effect an emergency plan that meets:

( l )

l2)

. ( 3)

Regulatory requirements of 10 CFR Part 50, Appendix E.

Regulatory Position Statements in Regulatory Guide 1.101 (March 1977)

  • The Ess~ntial Planning Elements in NUREG-75/111 and Supplement l thereto defined by NRR as significant for fuel load and low pov1er testing.
u.

In-Plant Radiation Monitoring - Partial (III.0.3.3)

PSE&G shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration throughout the plant under accident conditions.

v.

Communications (III.A.3.3)

PSE&G shall install and maintain direct dedicated telephone lines between the Salem plant control room, site technical support center and NRG Incident Response Center in Bethesda, Maryland.

(6)

Conditions on Operations Beyond Zero Power Testing The following conditions shall be completed to the satisfaction of the Commission prior to proceeding beyond zero pm*1er testing.

The following conditions are related to matters specified in the TMI Action Plan, and the Near Term Operating License (NTOL) requirements dated February 6, 1980, and are applicable to operation beyond zero power testing. Each of the following conditions references the appropriate section of Part II of Supplement No. 4 to the Safety Evaluation Report (NUREG-0517, Suppl. No. 4) for the Salem Generating Station, Unit 2 and follows the numbering sequence utilized in the February 6, 1980 NTOL Requirements list.

a.

Short-Term Effort - Analysis and Procedure Modifications (I.C.1)

PSE&G shall revise its emergency operating instructions for dealing with small break LOCAs and inadequate core cooling based on its analysis of these events and the vendor guidelines derived from these analyses.

These requirements supersede the requirements specified in Item 7(b) of IE Bulletin No.79-06A.

b.

NSSS Vendor Review of Low Power Test Procedures (I.C.7)

PSE&G's low power test procedures shall be reviewed by the nuclear steam supply system vendor, Westinghouse, and documentation of the review submitted to NRC.

c.

Low Power Test Program (I.G.)

As set forth in Section I.G., PSE&G shall obtain staff approval of a low power test program.

d.

Additional Accident Monitoring Instrumentation (II.F.1)

PSE&G shall comply with the requirements contained in NUREG-0578, Section 2. 1.8.b, as clarified in NRC letter to operating license applicants, dated November 9, 1979.

}

7 D.

PSE&G shall report any violations of the requirements* contained in Sections 2.C.(5) and (6) of this license in accordance with the prompt reporting requirements of Section 6.9.l.8 of Appendix A, which is attached to and is a part of this license.

E.

Prior to fuel loading, the licensee shall have fully implemented, to the satisfaction of the Commission, the final provisions of the physical security plan, withheld from public disclosure pursuant to 10 CFR Part 2.790(d), entitled "Salem Generating Station, Security Plan, Revision 4, Amendment 2 11 dated August 3, 1978.

The licensee shall maintain in effect the physical security plan, withheld from public disclosure pursuant to 10 CFR 2.790{d),

entitled, 11Salem Generating Station Security Plan, Revision 4, Amendment 2 11

, dated August 3, 1978, as amended in accordance with the provisions of 10 CFR 50.54(p).

In addition to all other commitments contained in the physical security plan, all keys, locks, combinations, and related equipment used to control access to protected and vital areas shall be controlled to reduce the probability of compromise.

Whenever there is evidence that any key, lock com bi nation, or related equiprrent may have been compromised, it shall be changed.

Upon termination of employment of any employee, keys, locks combinations, and related equipment to which that employee had access, shall be changed.

F.

This license is subject to the following condition for the protection of the environment:

Before engaging in additional construction or operational activities which may result in an en vi ronmenta l impact that was not evaluated by the Commission, the licensee will prepare and record an environmental evaluation of such activity. When the evaluation indicates that such activity may result in a significant adverse environmental impact that was not evaluated, or that is significantly greater

- 16 ':"

than that evaluated, in the Final Environ-mental Statement or any addendum thereto, the licensees shall provide a wr1tten evalua-tion of such activities and obtain prior approval from the Director, Office of Nuclear Reactor Regulation.

G.

This license is effective as of the date of issuance and shall expire one year after that date.

Attachment:

Appendices A and B Technical Specifications Date of Issuance:

FOR THE NUCLEAR REGULATORY COMMISSION Harold R. Denton, Director Office of Nuclear Reactor Regulation

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