Text

Forwards Util Response to NUREG-0737 TMI-related Requirements.Meeting to Discuss Response Requested
	ML18026A221
Person / Time
Site:	Susquehanna
Issue date:	01/22/1981
From:	Curtis N; PENNSYLVANIA POWER & LIGHT CO.
To:	Youngblood B; Office of Nuclear Reactor Regulation
References
	RTR-NUREG-0737, RTR-NUREG-737 PLA-614, NUDOCS 8101260370
	Download: ML18026A221 (104)
	v • d • e

TWO NORTH NINTH STREET, ALLENTOWN, PA. 18101 PHONEME (215) 770-5151 I

NORMAN W. CURTIS Vice President ~ Engineering tk Construction-Nuclear 770.5381 u A

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January 22, 1981 s> ".

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t Mr. B. J. Youngblood, Chief Licensing Projects Branch 1 Division of Project Management U.S. Nuclear Regulatory Commission Washington, D.C. 10555 SUSQUEHANNA STEAM ELECTRIC STATION RESPONSE TO TMI RELATED REQUIREMENTS Docket Nos. 50-387 ER 100450 PILE 841-12 PLA-614 and 50-388

Dear Mr. Youngblood:

Attached is the response to TMI related requirements for 'Susquehanna Steam Electric station (SSES) Units 1 and 2. We would like to arrange a meeting to discuss this document with the NRC. We will contact the Project Manager for SSES in the near future to establish a date. If you have any comments,

~

please call me.

.Very truly yours, N. W. Curtis Vice President-Engineering 6 Construction-Nuclear DPM/mks Attachment

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5 cc: R. M. Stark ~"";,'~~~gg~ ) QOCumeat RRG91 QQÃf DOCS F!LE i'ENNSYLVANIA POWER 8 .LIGHT COMPANY

RESPONSES TO TMI RELATED REQUIREMENTS TABLE OF CONTENTS PAGE X.O ORGANIZATION X.0-1'X.

X.l RESPONSE TO REQUIREMENTS IN NUREG 0737 1-1 X.l.l Shift Technical Advisor (I.A.l.1) X.l-l X.l.2 Shift Supervisor Responsibilities (I.A.l.2) X.1-3 K.l.3 Shift Manning (I.A.1.3) X.1-3

'.l.4 Immediate Upgrading Of Reactor Operator And X.1-6 Senior Reactor Operator Training And Qualifications (I.A.2.1)

X.1.5 Administration Of Training Programs (I.A.2.3) X.l-7 K.1.6 Revise Scope And Criteria For Licensing X.l-8 Examinations (I.A.3.1)

X.l.7 Evaluation Of Organization And Management (I.B.1.2) X.l-9 X.l.8 Short-Term Accident And Procedure Review (I.C.1) K.1-10 X.1.9 Shift Relief And Turnover Procedures (I.C.2) K.1-13 X.l,lo Shift Supervisor Responsibility (I.C.3) X.l-14 K.l.ll Control Room Access (I.C.4) X.1-14 X.1.12 Feedback Of Operating Experience (I.C.5) X.1-14 X.l.13 Verify Correct Performance Of Operating X.l-15 Activities (INC.6)

X.l..14 NSSS Vendor Review Of Procedures (I.C.7) X.1-16 X.1.15 Pilot Monitoring Of Selected Emergency Procedures X.1-16 For Near Term Operating Licenses (I.C.8)

X.l..16 Control Room Design Review (I.D.1) X.l-16 X.1.17 Plant Safety .Parameter Display Console (I.D.2) X. 1-'17 X.1.18 Training During I,ow-Power Testing (I.G.1) X.1-18 X.l.19 Reactor Coolant System Vents (I.B.l) X.l-18 X.1.20 Plant Shielding (II '.2) X.1-19 X.1.21 Post-Accident Sampling (II.B.3) X.l-27 X.l.22 Training For Mitigating Core Damage (II.B 4) X.1-30 X.l.23 Relief And Safety Valve Test Requirements (II.D.1) X.1-34 X.1.24 Safety/Relief Valve Position Indication (II.D.3) X.1-35 X.1.25 Auxiliary Feedwater System Evaluation (II.E.1.1) X.1-36 X.l.26 Auxiliary Feedwater System Initiation And Flow X.l-36 (II.E.1.2) .

X.l.27 Emergency Power For Pressurizer Heaters (ICE.3.1) X.1-36 X.1.28 Dedicated Hydrogen Penetrations (II.E.4.1) X.l-36 X.1.29 Containment Isolation Dependability (II.E.4.2) X.1-37 X.l.30 Accident-Monitoring Instrumentation (II.F.l) X.1-39 X.1.31 Instrument For Detection Of Inadequate X.1-44 Core Cooling (II.F.2)

. X.1.32 Emergency Power For Pressurizer Equipment (II.G.1) X.1-45 X.l.33 Review ESF Valves (II.K.1.5) X.1-45 X.1.34 Operability Status (II.K.1.10) X.l-45

'.l.'35 Trip 'Pressurizer Low-Level Coincident X.1-45 Signal Bistables (II.K.l.l7)

Table of Contents (cont'd)

PAGE X.l.36 Operator Training For Prompt. Manual Reactor X.1-45 Trip (II.K.1.20).

X.1.37 Automatic Safety Grade Anticipatory X.l-45 Reactor Trip (II.K.1.21)

X.l.38 Auxiliary Heat RemovalMystem Procedures (II.K.1.22) K.l-45 X.l.39 Reactor Vessel Level Procedures (II.K:1.23) X.1-45 X.l.40 Commission Orders On Babcock And Wilcox X.l-45 Plants (II.K.2)

X.1.41 Automatic Power-Operated Relief Valve X.1-46 Isolation System (II.K.3.1)

X.l.42 Report On Power-Operated Relief Valve Failures X.1-46 (II.K.3.2)

X.l.43 Reporting Safety/Relief Valve Failures And X. 1-.46 Challenges (II.K.3.3)

X.1.44 Automatic Trip Of Reactor Coolant Pumps During A X.l-46 LOCA (II.K.3.5)

X.1.45 Evaluation Of Power-Operated Relief Valve Opening K.1-46 Probability (II.K.3.7)

X.l.46 Proportional Integral Derivative Controller , X.l-46 Modification (II,K.3.9)

X.l.47 Proposed Anticipatory Trip Modificati'on (II.K.3.10) X.1-46 X.l.48 Power-Operated Relief Valve Failure Rate (II.K.3.11) X.l-46 X.1.49 Anticipatory Reactor'rip On Turbine Trip (II.K.3.12) X.1-46 X.1.50

~ ~ Separation Of High Pressure Coolant Injection And X.1-47 Reactor Cooling Isolation Cooling System Initiation Ievels (II;K.3.13)

X.l.51 Modify Break-Detection Logic To Prevent Spurious Isolation X.1-47 Of High Pressure Coolant Injection And Reactor Core Isolation Cooling (II.K.3.15)

X.1.52 Reduction Of Challenges And Failures Of Relief Valves (II.K.3.16)

X.1.53 Report On Outages Of Emergency Core Cooling X.1-49 Systems (II.K.3.17)

X.1.54 Modification Of Automatic Depressurization Syst: em X.1-50 Logic (II.K.3.18)

X.l.55 Restart Of Core Spray And Low Pressure Coolant Injection Systems (II.K.3.21)

K.1.56 Automatic Switchover Of Reactor Core Isolation X.1-51 Cooling System Suction (II.K.3.22)

X.l.57 Confirm Adequacy Of Space Cooling For High Pressure X.l-52 Coolant Injection And Reactor Core Isolation Cooling Systems (II.K.3.24)

X.l.58 Effect Of Loss Of Alternating-Current Power On X.1-52 Recirculation Pump Seals (II.K.3.25)

X.1.59 Provide A Common Reference Level For Vessel Level X.1-53 Instrumentation (II.K.3.27)

K.1.60 Verify qualification Of Accumulators On Automatic X.1-54 Depressurization System Valves (II.K.3.28)

Table of Contents (cont'd)

PAGE X.l.61 Revised Small-Break Loss Of Coolant Accident X.1-54 Methods (II.K.3.30)

X.l.62 Plant-Specific Calculations To Show Compliance X.l-56 With 10CFR Part 50.46 ('II.K.3.31)

X.l.63 Evaluation Of Anticipated Transients With Single Failure X.1-56 To Verify No Fuel Cladding Failure (II.K.3.44)

X.1.64 Evaluation Of Depressurization With Other Than The X.l"57 Automatic Depressurization System (II.K.3.45)

'.l.65 Michelsen Concerns (II.K.3.46) X.1-58 X.1.66 Emergency Preparedness - Short Term (III.A.1.1) X.1-58

. X.l.67 Upgrade Emergency Support Facilities (III.A.1.2) X. 1-58 "

X.l.68 Emergency Preparedness - I,ong Term (III.A.2) X. 1-59 X.1.69 Integrity. Of Systems Outside Containment Likely X.1-63 To Contain Radioactive Material (III AD.l.l)

X.1.70 Inplant'odine Radiation Monitoring (III.D.3.3) X.1-64 X.1.71 Control Room Habitability Requirements (III.D.3.4) X.l-66

Table of Contents (Cont'd)

PAGE X.2 RESPONSE TO REQUIREMENTS IN NUREG 0694 X.2"1 X.2.1 Shift Technical Advisor (I.A.l.l) X.2-1 X.2.2 Shift Supervisor Administrative Duties (I.A.1.2) ~

X.2-1 X.2.3 Shift Manning (I.A.1.3) X.2-1 X.2;4 Immediate Upgrading Of Operator And Senior X.2-1 Operator Training And Qualifications (I.A.2.1)

X.2.5 Revise Scope And Criteria For Licensing Examinations X.2-2 (I.A.3.1)

X.2.6 Evaluation Of Organization And Management Improvements X.2-2 Of Near-Term Operating License Applicants (I.B.1.2)

X.2.7 Short Term Accident Analysis And Procedure X.2-2 Revision (I.C.1)

X.2.8 Shift Relief And Turnover Procedures'(I.C.2) X.2-2 X.2.9 Shift Supervisor Responsibilities (I.C.3) X.2-3 X.2.10 Control Room Access (I.C.4) X.2-3 X.2.11 ~

Procedures For Feedback Of Operating Experience To X.2-4 Plant Staff (I.C.5)

X.2.12 NSSS Vendor Review Of Procedures (I.C.7) X.2-4 X.2.13 Pilot Monitoring Of Selected Emergency Procedures X.2-5 For Near-Term Operating License Applicants (I.C.8)

X.2.14 Control Room Design'(I.D.1) X.2-5 X.2.15 Training During Low Power Testing (I.G.l) X.2-5 X.2.16 Reactor Coolant System Vents (II.B.1) X.2-6 X.2.17 Plant Shielding (II.B.2) X.2-6 X.2.18 Post-Accident Sampling (II.B.3) X.2-6 X.2.19 Training For Mitigating Core Damage (II.B.4) X.2-6 X.2.20 Relief And Safety Valve Test Requirements (II.D.l) X.2-6 X.2.21 Relief And Safety Valve Position Indication (II.D.3) X.2-6 X.2.22 Containment Isolation Dependability (II.E.4.2) X.2-6 X.2.23 Additional Accident Monitoring Instrumentation (II.F.l) X.2-6 X.2.24 Core Cooling Instruments (II.F.2) 'nadequate X.2-7 X.2.25 Assurance Of Proper ESF Functioning (II.K.1.5) X.2"7 X.2.26 Safety Related System Operability Status (II.K.1.10) X.2-7 X.2.27 Trip Pressurizer Low-Ievel Coincident Signal X.2-7 Bistables (II.K.1.17)

X.2.28 Operator Training For Prompt Manual Reactor X.2-8 Trip (II.K.1.20)

X.2.29 Automatic Safety Grade Anticipatory Trip X.2-8 (II.K.1.21)

X.2.30 Auxiliary Heat Removal Systems Operating X.2-8 Procedures (II.K.1.22)

X.2.31 Reactor Level Instrumentation (II.K.1.23) X.2-8 X.2.32 Commission Orders On Babcock And Wilcox Plants (II.K.2) X.2-9 X.2.33 Reporting Requirements For Safety/Relief Valve Failure X.2-9 Or Challenges (II.K.3.3)

X.2.34 Proportional Integral Derivative Controller (II.K.3.9) X.2-9 X.2.35 Anticipatory Reactor Trip Modification (II.K.3.10) X.2-9 X.2.36 Power Operated Relief Valve Failure Rate (II.K.F 11) X.2-9

Table of Contents (Cont'd)

. PAGE X.2.37 Anticipatory Reactor Trip On Turbine Trip (II.K.3.12) X.2-9 X.2.38 Emergency Preparedness-Short Term (III.A.1.1) X.2-10 X.2.39 Upgrade Emergency Support Facilities (III.A,1,2) X.2-11 X.2.40 Primary Coolant Sources Outside Containment (III.D.l.l) X.2-11 X.2.41 Inplant Radiation Monitoring (III.D.3.3) . X.2-11 X.2.42 Control Room Habitability (III.D.3.4) X.2-11

TABIES Table Number X.l.3-1 Interim Required Shift Staffing X.l.20-1 Shielding And Access Study

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K.1.22-1 Training Criteria For Hitigating Coze Damage

'22D/mb

X. RESPONSES TO TMI RELATED REQUIREMENTS X.O ORGANIZATION This chapter contains a response for each TMI-related requirement. The chapter is divided into sections, containing the responses to all requirements for applicants for operating licenses issued in a single document. Consult the table of contents to identify what section provides the responses for a given document.

Each section addresses all the requirements in its corresponding document.

A response is only given to the most recent in the series of requirements which contains an explanatory text. For example, if an explanatory text of requirement I.A.l.l appears on both NUREG 0737 and NUREG 0694, a response is provided to NUREG 0737 since it supersedes all previous requirements.

If requirement I.A.1.2 appears in both NUREGs 0737 and 0694, but the only explanatory test is in NUREG 0694, the response is provided to NUREG 0694 utilizing the implementati'on dates of NUREG 0737.

1 X,o-l

X.l RESPONSE TO RE UIREHENTS IN NUREG 0737 X.l.l SHIFT TECHNICAL ADVISOR (I.A.l.l)

X.l.l.l Statement of'Re uirement Each licensee shall provide an on-shift technical advisor to the shift supervisor. The shift technical-advisor (STA) may serve more than one unit at a multiunit site if qualified to perform the advisor function for the various units.

The STA shall have a bachelor's degree or equivalent in a scientific or engineering discipline and have received specific training in the response and analysis of the plant for transients and accidents. The STA shall also receive training in plant design and layout, including the capabilities of instrumentation and controls in the control room. The licensee shall assign normal duties to the STAs that pertain to the engineering aspects of assuring safe operations of the plant, including the review and evaluation operating expe'rience.

'f The need for the STA position may be eliminated when the qualifications of the shift supervisors and senior operators have been upgraded and the interface in the control room has been acceptably upgraded. man-'achine However, until those long-term improvements are attained, the need for an STA program will continue.

The staff has not yet established the detailed elements of the acadamic and

,training requirements of the STA beyond the guidance given in the Vassallo letter on November 9, 1979. Nor has the staff made a decision on the level of upgrading required for licensed operating personnel and the man-machine interface in the control room that would be acceptable for eliminating the need of an STA. Until these requirements for eliminating the STA position have been established, the staff continues to require that, in addition to the staffing requirement specified in subsection X.1.3, an STA be available for duty on each operating shift when a plant is being operated in Hodes 1-3 for a BMR. At other times, an STA is not required to be on duty.

Since the November 9, 1979 letter was issued, several efforts have been made to establish, for the longer term, the minimum 1evel of experience, education, 'and training for STAs. These efforts include work on the revision to ANS-3.l,,work by the Institute of Nuclear Power Operations (INPO), and internal staff efforts.

INPO has made available a document entitled "Nuclear Power Plant Shift Technical Advisor--Recommendations for Position Description, Qualifications, Education and Training." Sections 5 and 6 of the INPO document describe the education, training, and experience requirements for STAs. The NRC staff finds that the descriptions as set forth in Sections 5

and 6 of Revision 0 to the INPO document are an acceptable approach for'he selection and training of personnel to staff the STA positions. (Note:

This should not be interpreted to mean that this is an NRC requirement at this time. The intent is to refer to the INPO document as acceptable for interim guidance for a utility in planning its STA program over the long term (i.e., beyond the January 1, 1981 requirement to have STAs in place in accordance with the qualification requirements specified in the staff's November 9, 1979 letter).)

Applicants for operating licenses. shall provide a description of their STA training and requalification program in their application, or amendments thereto, on a schedule consistent with the NRC licensing review schedule.

Applicants for operating licenses shall provide a description of their long-term STA program, including qualification, selection criteria, training, and possible phaseout. The description shall be provided in the application, or amendments thereto, on a schedule consistent with the NRC licensing review schedule. The description shall include a comparison of the long-term program with the above mentioned INPO document.

X.l.l.2 Inter retation Develop a training program in compliance with the November 9, 1979 letter and submit a description to the NRC. Provide STA coverage for all operating shifts. Candidates will complete a training program and pass a certification examination prior to assumption of duties. Develop a long-term program to maintain or phaseout STAs.

X.l.1.3

~ ~ ~ Statement of Res onset STA coverage will be provided for all operating shifts.. A requirement for this position is included in the Technical Specifications. The development will be completed as follows:

(1) Develop procedures defining the STA duties and interfaces.

(2) Complete training program.

(3) Pass certification examinations (4) STAs assume duties on assigned shifts.

(5) Evaluate the need for continuation of STA program.

A description of the training program will be submitted to the NRC prior to fuel load. 'All training will be completed and STAs will be ready for shift assignment prior to fuel load. A description of the long term STA program .

will be submitted to the NRC prior to fuel load.

X.1.2

~ ~ SHIFT SUPERVISOR RESPONSIBILITIES (I.A.1.2)

No requirement stated NUREG 0737. Refer to Subsection X.2.2 which contains the response to the requirement stated in NUREG 0694.

X.1.3 SHIFT liANNING (I.A.1.3)

X.l.3.1 Statement of Re uirement Applicants for operating licenses shall include in their administrative procedures (required by license conditions) provisions governing required shift staffing and movement of key individuals about the plant. These provisions are required to assure that qualified plant personnel to man the operational shifts are readily available in the event of an abnormal or emergency situation. Interim requirements for shift staffing are given in

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Table X.1 3-1.

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These administrative procedures shall also set forth a policy. The objective of this policy should be to operate the plant with the required staff and develop working schedules such that use of overtime is avoided, to the extent practicable, for the plant staff who perform safety-related functions (e.g., senior reactor operators, reactor operators, health physicists, auxiliary operators, ISC technicians and key'aintenance personnel).

The staff recognizes that there are diverse opinions on the amount of overtime that would be considered permissible and that there is a lack of hard data on the effects of overtime beyond the generally recognized normal

.8-hour working day, the effects of shift rotation, and other factors. NRC ~

has initiated studies in this area. ~ Until a firmer basis is developed on working hours, the administrative procedures shall include as an interim measure the following guidance, which generally follows that of IE Circular No. 80-02.

In the event that overtime must be used (excluding extended periods of shutdown for refueling, major maintenance or major plant modifications),

the following overtime restrictions should be followed:

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

(2) There should be a break of at least 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (which can include shift turnover time) between all work periods.

(3) An individual should not work more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7-day period.

(4) An individual should not be required to work more than 14 consecutive days without having 2 consecutive days off.

However, recognizing that circumstances may arise requiring deviation from the above restrictions, such deviation shall be authorized by the lant manager or his deputy, or higher levels of management in accordance with published procedures and with appropriate documentation of the cause.

If a reactor operator or senior reactor operator has been workini g more th an ours uring periods of extended shutdown (e.g:, at duties away from the control board), such individuals shall not be assigned shift duty in the control room without at least a 12-hour break preceding such an asignment.

NRC encourages the development of a staffing policy that would permit the licensed reactor operators and senior reactor operators to be eriodicall er duties away from the control board during their normal assi g ned too oother tours of duty.

If a reactor operator is required to work in excess of 8 continuous hours, he shall be periodically relieved of primary duties at the control board, such that periods of duty at the board do not exceed about 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months at a time The guidelines on overtime do not apply to the shift technical advisor provi ed he or she is provided sleeping accommodations and a 10-minute availability is assured.

Operating license applicants shall complete these administrative procedures before fuel load.

X.1.3.2~ Inter retation None required. ~

K.l.3.3 Statement of Res onse Th e facility staffing requirements are presented in Subsection 6.2.2 of the Technical Specifications. These requirements are consistent with those given in Table X.I.3-1.

A policy for working hours will be included in the administrative

'rocedures. This policy will be developed by fuel load. Appropriate documentation will be made available for .review by Region I ISE.

X.l-4

X.l.4 11BKDIATE UPGRADING OF REACTOR OPERATOR AND SENIOR REACTOR OPERATOR TRAINING AND UALIFICATIONS (I,A.2.1)

X.1.4.1 Statement of Re uirement Applicants for senior operator licenses shall have 4 years of responsible power plant experience: Responsible power plant experience should be that obtained as a control room operator (fossil or nuclear) or as a power plant staff engineer involved in the day-to-day activities of the facility, commencing with the final year of construction. A maximum of 2 years power plant experience may be fulfilled by academic or related technical training, on a one-for-one time basis. Two years shall be nuclear power plan't experience. At least 6 months of the nuclear power plant experience shall be at the plant for which he seeks a license. Effective date:

Applications received on or after >iay 1, 1980.

Applicants for senior operator licenses shall have held an operator's license for 1 year. Effective Date: Applications received after December 1, 1980. NRC has not imposed the 1-year experience requirement on cold applicants for SRO licenses. Cold applicants are to work on a facility not.

yet in operation; their training programs are designed to supply the equivalent of the experience not available to them.

Senior'perator'": Applicants shall have 3 months of shift training as an extra man on shift.

Control room operator-": Applicants shall have 3 months training on shift as an extra person in the control room. Effective date: Applications received after August 1, 1980.

Training programs shall be modified, as necessary, to provide:

1) Training in heat transfer, fluid flow and thermodynamics.

2) Training in the use of installed plant systems to control or mitigate an accident in which the core is severely damaged.

3) Increased emphasis on reactor and plant transients. Effective date: Present programs have been modified in response to Bulletins and Orders. Revised programs should be submitted for OLB review by August 1, 1980 applicants will be required to meet unique qualifications

':Precritical designed to accommodate the fact that their facility has not yet been in operation.

INTERIM REQUIRED SHIFf STAFFING One Unit, Two Units Two Units Three Units One Control One Control Two Control Two Control Operating Status Room Room Rooms Rooms One Unit Operating"" 1 SS (SRO) 1 SS (SRO) 1 SS (SRO) 1 SS (SRO) 1 SRO 1 SRO 1 SRO 1 SRO 2 RO 3 RO 3 RO 4 RO 2 AO 3 AO 3 AO 4 AO Two Units Operating"'A 1 SS (SRO) 1 SS (SRO) 1 SS (SRO) 1 SRO 2 SRO 2 SRO ) Only 1 SRO 8 4 ROs required 3 RO 4 RO 5 RO ) if both units are operated 3 AO 4 AO ) from one control room 5 AO All Units Operating='A 1 SS (SRO) 1 SS (SRO) SS (SRO) 1 SRO 2 SRO 2 SRO 3 RO 4 RO 5 RO 3 AO 4 AO' 5 AO All Units Shut Down 1 SS (SRO) 1 SS (SRO) 1 SS (SRO) 1 SS (SRO) 1 RO 2 RO 2 RO 3 RO 1 AO 3 AO 3 AO 5 AO SS - shift supervisor RO - licensed reactor operator SRO - licensed senior reactor operator AO auxiliary operator NOTE: (1) In order to operate or supervise the operation of more than one unit, an operator (SRO or RO) must hold an appropriate, current license for each such unit.

(2) In addition to the staffing requirements indicated in the table, a licensed senior operator will be required to directly supervise any core alteration activity.

(3) See item I.A.l.l for shift technical advisor requirements.

"'odes 1 through 3.

44K/cak

Certifications completed pursuant to Sections 55.10(a)(6) and 55.33a(4) and (5) of 10 CFR Part 55 shall be signed by the highest level of corporate management for plant operation (for example, Vice President for Operations). Effective date: Applications received on or after May 1, 1980

'.l.4.2 Inter retation, None required.

X.l.4.3 Statement of Res onse A program will be established to assure that all reactor operator and senior reactor operator license candidates (beyond the initial compliment required to startup Units 1 8 2) have the prescribed experience, qualifications, and training. The initial startup crews will have completed extensive training devised in part to recognize. the non-operational status of the units. This program includes real time training on a simulator which duplicates the actual unit and thus in many respects equates to the experience requirements. The program is described in Subsection 13.2.

X.l.5 ADMINISTRATION OF TRAINING PROGRAMS (I.A.2.3)

X.l.5.1 Statement of Re uirement Pending accreditation of training institutions, licensees and applicants for operating licenses will assure that training center and facility instructors who teach systems, integrated responses, transient, and

,simulator courses demonstrate senior reactor operatior (SRO) qualifications and be enrolled in appropriate requalification programs.

Training center and facility instructors who teach systems, integrated responses, transient and simulator courses shall demonstrate their .

competence to NRC by successful completion of a senior operator examination. Effective date: Applications should be submitted to later than August 1, 1980 for individuals who do not already hold a senior operator license.

Instructors shall be enrolled in appropriate requalification programs to assure they are cognizant of current operating history, problems, and changes to 'procedures and administrative limitations. Effective date:

Programs should be initiated May 1, 1980. Programs should be submitted to OLB for review by August 1, 1980.

X.1-7

X.1.5.2 Inte retation Instructors who teach systems specific to BWRs,'ntegrated responses, transients, and simulator courses will take and pass the examination for senior reactor operators (SRO).

X.l.5.3 Statement of Res onse A requalification program has been established for instructors. The above described instructors have either passed or are scheduled to take the SRO examination.

X.l.6 REVISE SCOPE AND CRITERIA FOR LICENSING EXAMINATIONS (I.A.3.1)

X.l.6.1 Statement of Re uirement A new category shall be added to the operator written examination entitled, "Principles of Heat Transfer and Fluid Mechanics."

, A new category shall be added to the senior operator written examination entitled, "Theory of Fluids and Thermodynamics."

Time limits shall be imposed for completion of the written examinations;

1. Operator: 9 hours.

2. Senior Operator: 7 hours.

The passing grade for the written examination shall be 80/ overall and 70$

in each category.

All applicants for senior operator licenses shall be required to be administered an operating test as well as the written examination.

Effective date: Examinations administered on or after May 1, 1980.

Applicants will grant permission to NRC to inform their facility management regarding the results of the examinations for purposes of enrollment in requalification programs. Applications received on or after May 1, 1980.

'I Simulator examinations will be included as part of the licensing examinations.

X.l.6.2 Inter retation None required

X.l.6.3 Statement of Res onse Reactor Operator and Senior Reactor Operator training has been upgraded to include the subject material described in this requirement. Refer to Subsection X.l.4.3 for the response to requirement I.A.2.1, "Immediate Upgrading of Reactor Operator and Senior Reactor Operator Training and

(}ualifications." Candidates will be prepared to take and pass NRC exams based on the new criteria. The SSES simulator will be available for the simulator portion of exams. Application packages will include a release which permits the NRC to inform PPRL management of exam results.

X.l.7 EVALUATION OF ORGANIZATION AND HANAGEMENT (I.B.1.2)

X.1.7.1 Statement of Re uirement Each applicant for an operating license shall establish an onsite independent safety engineering group (ISEG) to perform independent reviews of plant operations.

The principal fun'ction of the ISEG is to examine plant operating characteristics, NRC issuances, Licensing Information Service advisories; and other appropriate sources of plant design and operating experience information that may indicate areas for improving plant safety. The ISEG is to perform independent review and audits of plant activities including maintenance, modifications, operational problems, and operational analysis, and aid in the establishment of programmatic requirements for plant activities. Where useful improvements can be achieved, it is expected that this group will develop and present detailed recommendations to corporate

.management for such things as revised procedures or equipment modifications.

Another function of the ISEG is to maintain surveillance of plant operations and maintenance activities to provide independent verification that these activities are performed correctly and that human errors are reduced as far as practicable. The ISEG will then be in a position to advise utility management on the overall quality and safety of operations.

The ISEG need not perform detailed audits of plant operations and shall not be responsible for sign-off functions such that it becomes, involved in he operating organization.

The new ISEG shall not replace the plant operations review committee (PORC) and the utility's independent review and audit group as'specified by current staff guidelines (Standard Review Plan, Regulatory. Guide 1.33, Standard Technical Specifications). Rather, it is an additional independent group of a minimum of five dedicated, full-time engineers, located onsite, but reporting offsite to a corporate official who holds a high-level, technically oriented position that is not in the management chain for power production. The ISEG will increase the available technical expertise located onsite and will provide continuing, systematic, and K.l-9

independent assessment of plant activities, Integrating the shift technical advisors (STAs) into the ISEG in some way would be desirable in that it could enhance the group's contact with 'and knowledge of day-to-day plant operations and provide additional expertise. However, the STA on shift is necessarily a member of the operating staff and cannot be independent of it.

It is expected that the ISEG may interface with the quality assurance (gA) organization, but preferably should not be an integral part of the QA organization.

The functions of the ISEG require daily contact with the operating personnel and continued access to plant facilities and records. The ISEG review functions can, therefore, best be carried out by a group physically located onsite. However, for utilities with multiple sites, it may be possible to perform portions of the independent safety assessment function in a centralized location for all the utility's plants. In such cases, an onsite group still is required, but it may be slightly smaller than would be the case if it were performing the entire independent safety assessment function. Such cases will be reviewed on a case-by-'case basis.

This requirement shall be implemented prior to issuance of an operating license.

Refer to Subsection X.2.6 for the response to additional requirements contained in NUREG 0694.

X.l.7.2 Inter retation None required.

X.1.7.3 Statement of Res onse

/

PPSL has established a Nuclear Safety Assessment Group which will perform all functions of the ISEG. This group is described in a letter from N.W.

Curtis to B. J. Youngblood on December 8, 1980 (PLA-585).

X.l.8 SHORT-TEE1 ACCIDENT AND PROCEDURE REVIEW (I.C.1)

X.l.8.1 Statement of Re uirement Reanalysis of small break LOCAs, transients, accidents, and inadequate core cooling and preparation of guidelines for development of emergency procedures should be completed and submitted to the NRC for review by January 1, 1981. The NRC staff will review the analyses and guidelines and determine their acceptability by July 1, 1981, and will issue guidance to licensees on preparing emergency procedures from the guidelines. Following NRC approval of the guidelines, licensees and applicants for operating licenses issued prior to January 1, 1982, should revise and implement their emergency procedures at the first refueling outage after January 1, 1982.

Applicants for operating licenses issued after January 1, 1982 should implement the procedures prior to operation. This schedule supersedes the implementation schedule included in NUREG-0578, Recommendation 2.1.9 for item I.C.1(a)3, Reanalysis of Transients and Accidents. For those licensees and/or owners groups that will have difficulty in attaining the January 1, 1981 due date for submittal of guidelines, a comprehensive program plan, proposed schedule, and a detailed justification for all delays and problems shall be submitted in lieu of the guidelines.

X.1.8.2 Inter retation Utilize the BWR Owners'roup guidelines to develop emergency procedures for accidents and transients. These guidelines should also be used as an input to the operator training program.

X.1.8.3 Statement of Res onse In the Clarification of the NUREG-0737 requirement "for reanalysis of transients and accidents and inadequate core cooling and preparation of guidelines for development of emergency procedures," NUREG-0737 states:

Owners'roup or vendor submittals may be referenced as appropriate'o support this reanalysis. If owners'roup or vendor submittals have already been forwarded to the staff for review, a brief description of the submittals and justification of their adequacy to support guideline development is all that is required.

PPM has participated, and will continue to participate, in the BWR Owners' Group program to develop Emergency Procedure Guidelines for General

~ Electric Boiling Water Reactors. Following are a brief description of the submittals to date, and a justification of their adequacy to support guideline development.

A. Descri tion of Submittals (1) NED0-24708, "Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors," August, 1979; including additional sections submitted in pre-publication form since August, 1979.

(a) Section 3.1.1 (Small Break IOCA).

Description and analysis of small break loss-of-coolant events, considering a range of break sizes,.location,,and conditions, including equipment failures and operator errors; description and justification of analysis methods.

(b) Section 3.2.1 (Loss of Feedwater) - revised and resubmitted in prepublication form March 31, 1980.

X.1-11

Description and analysis of loss of feedwater events, including cases involving stuck-open relief valves, and including equipment failures and operator errors; description and justification of analysis methods.

(c) Section 3.2.2 (Other Operation'al Transients) - submitted in prepublication form March 31, 1980; revised and resubmitted in prepublication form August 22, 1980.

Description and arralysis of each FSAR Chapter 15 event resulting in a reactor system transient; demonstration of applicability of analyses of Sections 3.1.1, 3.2.1, and 3.5.2.1 to each event; demonstration of applicability of Emergency Procedure Guidelines to each event.

k (d) Section 3.3 (BWR Natural and Forced Circulation)

Description of natural and forced circulation cooling; factors influencing natural circulation, including noncondensibles; reestablishment of forced circulation under transient and accident conditions.

(e} Section 3.5.2.1 (Analyses to Demonstrate Adequate Core Cooling) - submitted in prepublication form November 30, 1979; revised and resubmitted in prepublication form September 16, 1980.

Description and analysis of loss-of-coolant events, loss of feedwater events, and stuck-open relief valve events, including severe multiple equipment failures and operator errors which, if not mitigated, could result in conditions of inadequate core cooling.

Section 3.5.2.3 (Diverse Methods of Detecing Ade'quate Core Cooling) - submitted in prepublication form December 28, 1979.

Description of indications available to the BWR operator for the detection of adequate core cooling (detailed instrument rsponses are described in Secti'ons 3.1.1, 3.2.1, and 3.5.2 ').

(g) Section 3.5.2.4 (Justification of Analysis Methods)-

submitted in prepublication form September 16, 1980.

Description and .justification of analysis methods for extremely degraded cases treated in Section 3.5.2.1.

(2) BWR Emergency Procedure Guidelines (Revision 0) - submitted in prepublication form June 30, 1980.

Guidelines for BWR Emergency Procedures based on identification and response to plant symptoms; including a range of equipment failures and operator errors; including severe multiple equipment failures and operator errors which, if not mitigated, would result in conditions of inadequate core cooling; including conditions when core cooling status is uncertain or unknown..

B. Ade uac of Submittals The submittals described in~aragraph A have been discussed and reviewed extensively among the BWR Owners'roup, the General Electric Company, and the NRC staff. The NRC staff has. found (NUREG-.0737, page I.C.1-3) that "the analysis and guidelines submitted by the General Electric Company (GE) Owners'roup...comply with the requirements (of the NUREG-0737 clarification)." In Reference 1, the Director of the Division of Licensing states, "we find the Emergency Procedure Guidelines acceptable for trial implementation (on six plants with applications for operating licenses pending)."

PPSL believes that in view of these findings, no further detailed justificatio'n of the analyses or guidelines is necessary at this time.

Reference 1 further states, "(during the course of implementation we may identify areas that require modification or further analysis and justification." The enclosure to Reference 1 identifies several PPGL will work with the BWR Owners'roup. in responding to such'reas.

such requests.

By our commitment to work with the Owners'. Group on such requests, on schedules mutually agreed to by the NRC and the Owners'roup, and by reference to the BWR Owners'roup analyses and guidelines already i submitted, our response to the NUREG-0737 requirement "for reanalyses of transients and accidents and inadequate core cooling and preparation of guidelines for development of emergency procedures" by January 1, 1981, is complete.

Emergency procedures based on these guidelines will be developed and available for use by fuel load. Appropriate documentation vill be made available for review by Region I ISE.

References (1) Letter, D. G. Eisenhut (NRC) to S.T. Rogers (BWR Owners'roup),

regarding Emergency Procedure Guidelines, October 21, 1980.

K.l.9 SHIFT RELIEF AND TURNOVER PROCEDURES (I.C.2)

No requirement stated in NUREG 0737. Refer to Subsection. X.2:8 which contains the response to the requirement in NUREG 0694.

X.l.10

~ ~ SHIFT SUPERVISOR RESPONSIBILITY (I.C.3)

No requirement stated in NUREG 0737. Refer to subsection X.2.9 which contains the response to the requirement in NUREG 0694.

X.1.11 CONTROL ROO>1 ACCESS (I.C.4)

No requirement stated in NUREG 0737. Refer to Subsection X.2.10 which contains the response to the requirement in NUREG 0694.

X.l.12 FEEDBACK OF OPERATING EXPERIENCE (I.C.5)

X.l.12.1 Statement of Re uirement Applicants for an operating license shall prepare procedures to assure that information pertinent to plant safety originating inside or outside the utility organization is continually supplied to operators and other personnel and is incorporated into training and retraining programs. These procedures shall:

(1) .Clearly identify organizational responsibilities for review of operating experience, the feedback of pertinent information to operators and other personnel, and the incorporation of such information into training and retraining programs; (2) Identify the administrative and technical review steps necessary in translating recommendations by the operating experience assessment group into plant actions (e.g., changes to procedures, operating orders);

(3) Identify the recipients of various categories of information from operating experience (i.e., supervisory personnel, shift technical advisors, operators, maintenance personnel, health physics technicians) or otherwise provide means through which such information can be readily related to the job functions of the recipients; (4) Provide means to assure that affected personnel become aware of and understand information of sufficient importance that should not wait for emphasis through routine training and retraining programs; (5) Assure that plant personnel do not routinely receive extraneous and unimportant information on operating experience in such volume that it would obscure priority information or otherwise proficiency; detract from overall job performance and

(6) Provide suitable checks to assure that conflicting or contradictory information is conveyed to operators and other personnel until resolution is reached; and, (7) Provide periodic internal audit to assure that the feedback program functions effectively at all levels.

This requirement, shall be implemented prior to issuance of an operating license.

K.1.12.2 Inter retation None required.

X.1.12.3 Statement of Res onse The Shift Technical Advisor (STA) will be the focal point for dissemination of operating experience information to the operations section. Procedures will be developed defining this function and the interfaces among the STAs .

and the Nuclear Safety Assessment Group, Nuclear Training, Operations, and Industry Events Review Program (IERP).

These procedures will be completed and implemented prior to receiving a fuel load license. After completion, appropriate documentation will be made available for review by Region I ISE.

X '.13 VERIFY CORRECT PERFORMANCE OF OPERATING ACTIVITIES (I.C,6)

.X.l.13.1 Statement of

~ ~ ~ Re uirement Licensees'rocedures shall be reviewed and revised, as necessary, to assure that an effective system of verifying the correct performance of operating activities is provided as a means of reducing human'rrors and improving the quality of normal operations. This will reduce the frequency of occurrence of situations that could result in or contribute to accidents. Such a verification system may include automatic system status monitoring, human verification of operations and maintenance activities independent of the people performing the activity (see NUREG-0585, Recommendation 5), or both.

Implementation of automatic status monitoring if required will reduce the extent of human verification of operations and maintenance activities but will not eliminate the need for such verification in all instances. The procedures adopted by the licensees may consist of two phases--one before and one after installation of automatic status monitoring equipment, if required, in accordance with item I.D.3.

Procedures must be reviewed and revised prior to fuel load.

e X.l-15

X.l.13.2 Inte retation None required.

X.1.13.3 Statement of Res onse Administrative'ontrols will be developed by fuel load to provide verification of correct performance of surveillance and maintenance activities on safety-related equipment. These controls will utilize automatic status monito'ring (control room indications) presently available and operability testing whre appropriate. Independent verification by a second qualified person will be employed as applicable. These administrative controls will provide shift supervision with control and knowledge of equipment status. Procedures being written in response to requirements II.K.1.5 and II.K.1.10 (see Subsections X.2.25 'and X.2.26) will be coordinated with these administrative controls. Appropriate documentation will be made available for review by Region I IGE.

X.l.14 NSSS VENDOR REVIEW OF PROCEDURES (I.C.7)

No requirement stated in NUREG 0737. Refer to Subsection'.2,12 which contains the response to the requirement in NUREG 0694, X.1.15 PILOT MONITORING OF SELECTED EMERGENCX PROCEDURES FOR NEAR TERM OPERATING LICENSES (I.O.8)

No requirement stated in NUREG 0737. Refer to Subsection X.2.13 which contains the response to the requirement in NUREG 0694.

'I X.1.16 CONTROL ROOM DESIGN REVIEW (I.D.l)

X.1.16.1 Statement of Re uirement All licensees and applicants for operating licenses will be required to conduct a detailed control-room design review to identify and correct design deficiencies. This detailed control-room design review is expected to take about a year. Therefore, the Office of Nuclear Reactor Regulation (NRR) requires that those applicants for operating licenses who are unable to complete this review prior to issuance of a license make preliminary assessments of their control rooms to identify significnat human factors and instrumentation problems and establish a schedule (to be approved by NRC) for correcting deficiencies. These applicants will be required to complete the more detailed control room reviews on the same schedule as licensees with operating plants.

Applicants will find it of value to refer to the draft document NUREG/CR-1580, ."Human Engineering Guide to Control Room Evaluation," in performing the preliminary assessment. NRR will evaluate the applicants preliminary

assessments including the performance by NRR of onsite review/audit. The NRR onsite review/audit will be on a schedule consistent with licensing needs, This requirement shall be met prior to fuel load.

X.1.16.2 Inter retation Applicants for operating licenses are required to perform a preliminary control room design assessment which should be based on NUREG/CR-1580.

This assessment will be reviewed by the NRC, who will subsequently recommend changes for correcting deficiences. Applicants must submit for NRC approval a schedule for correcting these deficiencies.

Applicants will be required to perform a detailed control room design assessment following NUREG 0700 issuance. This assessment is not required to be completed prior to issuance of an operating license.

X.1.16.3 Statement of Res onse A preliminary design assessment of the SSES control room was performed and discussed with the NRC the week of October 27, 1980. This assessment is based on NVREG/CR-1580 as suggested above. The NRC has completed an

~

independent review of the control room. ,PP5L is currently waiting for the NRC comments. The preliminary assessment will be submitted to the NRC prior to fuel load.

X.1.17 PLANT SAFETY PARAhKTER DISPLAY CONSOLE I;D.2 X:1.17.1

~ ~ Statement of Re uirement Each applicant and licensee shall install a safety parameter display system (SPDS) that will display to operating personnel a minimum set of parameters which define the safety status of the plant. This can be attained through continuous indication of direct and derived variables as necessary to assess plant safety status.

The implementation date will be announced >>ith the issuance of NUREG-0696.

X.l.17 '2 Inter retation

~

None required.

. X.1.17.3 Statement of Res onse The response to this requirement will be incorporated into Appendix I of the Emergency Plan following issuance of NUREG 0696.

X.l.18 TRAINING DURING LOW-POWER TESTING (I.G.l)

No requirement stated in NUREG 0737. Refer to Subsection X.2.15 which contains the response to the requirement in NUREG 0694.

X.1.19 REACTOR COOLANT SYSTEM VENTS (II.B.l)

X.l.19.1 Statement of Re uirement Each applicant and licensee shall install reactor coolant system (RCS) and reactor pressure vessel (RPV) head high point vents remotely operated from the control room. Although the purpose of the system is to vent noncondensible gases from the RCS which may inhibit core cooling during natural circulation, the vents must not lead to an unacceptable increase in the probability of a loss-of-coolant accident (LOCA) or a challenge to containment integrity. Since these vents form a part of the reactor coolant pressure boundary, the design of the events shall conform to the requirements of Appendix A to 10 CFR Part 50, "General Design Criteria."

The vent system shall be designed with sufficient redundancy that assures a low probability of inadvertent or irreversible actuation.

Each licensee shall provide the following information concerning the design and operation of the high point vent system:

(1) Submit a description of the design, location, size, and power supply for the vent system along with results of analyses for loss-of-coolant accidents initiated by a break in the vent pipe.

The results of the analyses should demonstrate compliance with the acceptance criteria of 10 CFR 50.46.'2)

Submit procedures and supporting analysis for operator use of the vents that also include the information available to the operator for initiating or terminating vent usage.

Documentation shall be submitted by July 1, 1981. Modifications shall be completed by.l, 1982.

X.1.19.2 Inter retation None required.

X.1.19.3 Statement Of Res onse Present SSES design of reactor coolant and reactor vessel vent systems meet the requirements of the NRC.

The RPV is equipped with various redundant means to vent the reactor during all- modes of operation. All the valves involved are safety grade, powered

by essential busses and are capable of remote manual operation from the control room.

The largest portion of non-condensables are vented through sixteen (16) safety relief valves (PSV 141F013A-S) mounted on the-main steam lines.

These power operated relief valves satisfy the intent of the NRC position.

Information regarding the design, qualification, power source of these valves has been provided in Sections 5.1, 5.2.2, 6.2, 6.3, 7.3 and 15.

In addition to power operated relief valves, the RPV is equipped with various other means of high point venting. These are:

1. Normally closed RPV head vent valves (HV141-F001 and F002),

operable from control room which discharges to drywell equipment drain tank. (Subsection 5.1 and Figure 5.1-3a).

2. Normally open reactor head vent line 2 DBA-112 which discharges to main steam line "A". (Subsection 5.1 and Figure 5.1-3a).

3. Main steam driven RCIC and HPCI system turbines, operable from'he control room which exhaust to suppression pool. (Subsections 5.3, 6.3 and Figures 5.4-9a, 6.3-1a).

Although the power operated relief valves fully satisfy the intent of the NRC requirement these other means also provide protection against accumulation of non-condensables in the RPV.

The design of the RCS and RPV vent systems is in agreement with the generic capabilities proposed by the BWR Owners'roup, with the exception of isolation condensers. SSES is not equipped with isolation condensers.

X.1.20 Plant Shieldin (II.B.2)

X.1.20.1 Statement of Re uirement With the assumption of a postaccident release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalent of 50% of the core radioiodine, 100% of the core noble gas inventory, and 1%

of the core solids are contained in the primary coolant), each licensee shall perform a radiation and shielding-design review of the spaces around systems that may,.as a result of an accident, contain highly radioactive materials. The design review should identify the location of vital areas and equipment, such as the control room, radwaste control stations, emergency power supplies, motor control centers, and instrument areas, in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during postaccident operations of these systems'

Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or postaccident procedural controls. The design review shall determine which types of corrective actions are needed for vital areas thorughout the facility.

K.1.20.1.1 'ocumentation Re uired for Vital Area Access For vital area access, operating license applicants need to provide a summary of the shielding design review, a description of the review results, and a description of the modifications made or to be made to implement the result of the review. Also to be provided by the licensee:

(1) Source terms used including time after shutdown that was assumed for source terms in systems.

(2) Systems assumed to contain high levels of activity in a post-accident situation and jusitification for excluding any of those given in the "Clarification" of NUREG 0737.

(3) Areas assumed vital for post-accident operations including justification for exclusion of any of those given'n the "Clarification" of NUREG 0737.

(4) Projected doses to individuals for necessary occupancy times in vital areas and a dose rate map for potentially occupied areas.

X.l.20.1.2 Documentation Re uired for E ui ment uglification II.B.2 states, "Provide the information requested by the Commission Memorandum and Order on equipment qualification (CLI-80-21)." This memorandum, with regard to equipment qualification, requests information on environmental qualification of safety related electrical equipment.

X.l.20.2 Inter retation X.1.20.2.1 Source Terms The source term for recirculated depressurized coolant need not be assumed to contain noble gases, therefore the RHR shutdown cooling system which may initiate at low reactor pressure only will be assumed to contain solely halogens and particulates. The HPCI and LPCI systems do not recirculate reactor coolant but, rather, suppression pool water. They will also be essentially void of noble gases.

Leakage from systems outside of containment need not be considered as potential sources. Also, containment and equipment leakage (from systems outside containment) need not be considered as potential airborne sources within the reactor building. It follows that airborne sources and any X.1-20

other uncontained sources in the reactor building need not be considered in this 'shielding review.

X.1.20.2.2 Post-Accident S stems The standby gas treatment system, or equivalent, is given as a system which may contain high levels of radioactivity after an accident. Airborne activity from leakage of equipment outside containment has been clearly established as being outside the review requirements. Drywell leakage must then provide the activity processed by the SGTS. This review will 'assume the drywell does indeed leak to the reactor building to provide a source within the SGTS. However, this airborne source will not be evaluated any further in the review.

X.1.20.2.3 E ui ment uglification Provide a description of the environmental qualification program and results for safety related electrical equipment both inside and outside of containment. It is our understanding that radiation qualification of non-electrical safety related equipment need not be reported.

X.l.20.3 Statement of Res onse The'equired post-accident study will be divided into'wo parts; one dealing with a summary of the shielding design review plus vital area access, another dealing with equipment qualification. Detailed outlines of both parts are given in Tables X.l.20-1 and X.l.20-2.

A summary of the shielding design review and results will be submitted to the NRC by March 1981. A description of the methodology used to determine radiation doses to safety-related equipment will be submitted by March 1981. The results of the equipment qualification program is scheduled to be submitted in April 1981 in revision 2 of the SSES Environmental

(}ualification Report for Class lE Equipment.

X.l-21

TABLE X.1.20-1 SHIELDING AND ACCESS STUDY I. INTRODUCTION A. Background B. Intent of Review II. DESIGN REVIEW BASES A. Post-Accident Systems

l. Core Spray (CS)

2. High Pressure Coolant Injection (HPCI) 3, Low Pressure Coolant Injection (LPCI)

4. Reactor Core Isolation Cooling (RCIC)

5. Residual Heat Removal System (RHR) a) Shutdown cooling mode b) Suppression pool cooling mode c) Containment spray mode

6. Main Steam Isolation Valve - Leakage Collection System (MSIV-LCS)

7. Sampling Systems a) Existing reactor sampling system b) Containment atmosphere monitoring system c) Plant vent sampling station d) Post-accident sampling station

8. Standby Gas Treatment System - SGTS B. Radiation Source Release Fractions

1. Source A: Containment Atmosphere a) 100% noble gases b) 25/ halogens

2. Source B: Reactor liquids a) 100/ noble gases b) 50/ halogens c) 1/ particulates X.1-22

3. Source C: Suppression pool liquid a) 50/ halogens b) 1/ particulates

4. Source D: Reactor Steam a) 100$ noble gases b) 25/ halogens C. Source Term quantification

1. Decay time a) Assume instantaneous release

2. Dilution volumes a) Source A

1) Drywell free volume

2) Suppression pool free volume b) Source B

1) Reactor coolant system normal liquid volume c) Source C

1) Reactor coolant system volume

2) Suppression pool volume d) Source D

1) Reactor steam volume

3. Depletion Factors a) Credit taken for reactor steam activity depletion during RCIC system operation b) Credit taken for reactor steam activity depletion during HPCI system operation D. System/Source Summary X.1-23

E. Dose Integration Factors

1. Personnel radiation doses a) Exposures based on one year occupancy b) Occupancy factors

2. Equipment radiation doses a) 'Exposures +ased on one year III'. SHIELDING REVIEW METHODOLOGY A. Radiation dose calculation model

l. Employ point kernel shielding technique

2. Exclude small piping

3. Exclude shine over partial walls B. Post-accident radiation zone maps Exclude normal operating dose rates 2~ Exclude airborne dose rates C. Personnel Radiation Exposure Guidelines

1. 10CFR50, Appendix A, GDC 19 a) 5 Rem maximum exposure D, Vital Area Identification and Access Vital area clarification a) Vital areas appropriate to Susquehanna b) Vital areas not appropriate to Susquehanna c) Integrated doses to personnel occupying vital areas

2. Vital area access a) Basis b) Results IV. RESULTS A. Radioactive Decay. Effects

l. Apply radiation zone maps for times other than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months post-accident

B. Integrated Personnel Exposures

l. Occupancy factors

2. Frequency and duration factors for areas not continuously occupied.

Personnel doses C. Reactor Building Accessibility D. Control Room Accessibility X.l-25

TABLE X.1.20-2 E(}UIP1'IENT UALIFICATION I. INTRODUCTION A. Background B. Intent of Review II. QUALIFICATION METHODOLOGY-A. Establish Integrated Dose Zone Haps

l. Employ Chapter I Criteria a) Highest contact dose for equipment within shielded

'compartments b) Attenuated dose for equipment outside shielded compartments

2. Determine equipment not qualified B. Establish Dose versus Distance Curves

l. Apply to equipment within shielded compartments

2. Determine equipment not qualified C. Other Case by Case gualificaiton Procedures

l. Determine actual time required for operation

2. Determine decay time in transfer of activity

3. Evaluate repositioning of electrical- component III. RESULTS A. Complete Response in Reply to IE Bulletin 79-01B, Supplement 2 B. Equipment Not (}ualified by Above 'Procedures

1. Contact vendor for requalification

2. Replace X.l-26

K.1.21 POST-ACCIDENT SAMPLING (II.B.3)

X.l.21.1 Statement of Re uirement A design.and operational review of the reactor coolant and containment atmosphere sampling line systems shall be performed to determine the capability of personnel to promptly obtain (less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ) a sample under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 18-3/4 rem to the whole body or extremities, respectively. Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products. If the review indicates that personnel could not promptly and safely obtain the samples, additional design features or shielding should be provided to meet the criteria.

A design and operational review of the radiological spectrum analysis facilities'shall be performed to determine the capability to promptly quantify (in less than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) certain radionuclides that are indicators of the degree of core damage. Such radionuclides are noble gases (which indicate cladding failure), iodines and ces'iums (which indicate high fuel temperatures), and nonvolatile isotopes (which indicate fuel melting). The initial reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4 release. The review should also consider the effects of direct radiation from piping and components in the auxliary building and possible contamination and direct radiation from airborne effluents. If the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications for equipment procurement shall be undertaken to meet the criteria.

In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions. Procedures shall be provided to perform boron and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source term). Both analyses shall be capable of being completed promptly (i.e., the boron sample analysis within an hour and the chloride sample analysis .within a shift).

The following items are clarifications of requirements identified in NUREG-0578, NUREG-0660, or the September 13 and October 30, 1979 clarification

letters, (1) The licensee shall have the capability to promptly obtain reactor coolant samples and containment atmosphere samples. The combined time allotted for sampling and analysis should be 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months or less from the time a decision is made to take a sample.

(2) The licensee shall establish an onsite radiological and chemical analysis capability to provide, within the 3-hour time frame established above, quantification of the following:

X.1-27

(a) certain radionuclides in the reactor coolant and containment atmosphere that may be indicators of the degree of core damage (e.g., noble gases; iodines.and cesiums, and nonvolatile. isotopes);

(b) hydrogen levels in the containment atmosphere; (c) dissolved gases (e.g., Hp), chloride (time allotted for analysis subject to discussion below), and boron concentration of liquids.

(d) Alternatively, have inline monitoring capabilities to perform all or part of the above analyses.

Reactor coolant and containment atmosphere sampling during postaccident conditions shall not require an isolated auxiliary system (e.g., the letdown system, reactor water cleanup system to be placed in operation in order to use the sampling system.

Pressurized licnsee reactor coolant samples are not required if the can quantify the amount of dissolved gases with unpressurized reactor coolant samples; The measurement of either total dissolved gases or H 2 gas in reactor coolant samples is considered adequate. Heasuring the 02 concentration is recommended, but is not mandatory.

The time for a chloride analysis to be performed is dependent upon two factors: (a) if the plant's coolant water is seawater or brackish water and (b) if there is only a single barrier between primary containment systems and the cooling water. Under both of the above conditions the licensee shall provide for,a chloride analysis within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the sample being taken. For all other cases, the licensee shall provide for the analysis to be completed within 4 days. The chloride analysis does not have to be done onsite.

The design basis for plant equipment for reactor coolant and containment atmosphere sampling and analysis must assume that it is possible to obtain and analyze a sample without radiation exposures to any individual exceeding the criteria of GDC 19 (Appendix A, 10 CFR Part 50)(i.e., 5 rem whole body, 75 rem extremities). (Note that the design and operational review criterion was changed from the operational limits of 10 CFR Part 20 (NUREG-0578) to the GDC 19 criterion (October 30, 1979 letter from H.R. Denton to 'all licensees).)

The analysis of primary coolant samples for boron is required for PMRs. (Note that Revision 2 of Regulatory Guide 1.97, when

-issued, will likely specify the need for primary coolant boron analysis capability at BWR plants.)

X.l-28

If inline monitoring is used for any sampling and analytical capability specified herein, the licensee shall provide backup sampling through grab samples, and shall demonstrate the capability of analyzing the samples. Established planning for analysis at offsite facilities is acceptable. Equipment provided for backup sampling shall be capable of providing at least one sample per'ay for 7 days following onset of the accident and at least one sample per week until the accident condition no longer exists.

(9) The licensee's radiological and chemical sample analysis capability shall include provisions to:

(a) Identify and quantify the isotopes of the nuclide categories discussed above to levels corresponding to the source terms given in Regulatory Guide 1.3 or 1.4 and 1.7. Where necessary and practicable, the ability to dilute samples to provide capability for measurement and reduction of personnel exposure should be'provided. Sensitivity of onsite liquid sample analysis capability should be such as to permit measurement of nuclide concentration in the range from approximately 1 u Ci/g to 10 Ci/g.

(b) Restrict background levels of radiation in the radiological and chemical analysis facility from sources such that the sample analysis will provide results with an acceptably small error (approximately a factor of 2). This can be accomplished through the use of sufficient shielding around samples and outside sources, and by the use of ventilation system design which will control the presence of airborne radioactivity.

(10) Accuracy,,range, and sensitivity shall be adequate to provide pertinent data to the operator in order to describe radiological and chemical status of the reactor coolant systems.

(ll) In the design of the postaccident sampling and analysis capability, consideration should be given to the following items:

I (a) Provisions for purging sample lines, for reducing plateout in sample lines, for minimizing sample loss or distortion, for preventing blockage of sample lines by loose material in the RCS or containment, for appropriate disposal of the samples, and for flow restrictions to limit reactor coolant loss from a rupture of the sample line. The postaccident reactor coolant and containment atmosphere following a transient or accident. The sample lines should be as short as possible to minimize the volume of fluid to be taken -from containment. The residues of sample collection should be returned to containment or to a closed system.

X.1-29

(b) The ventilation exhaust. from the sampling station sho'uld be filtered with charcoal adsorbers and high-efficiency particulate air (HEPA) filters.

Operating License Applicants--Provide a description of the implementation of the position and clarification including PRIDs, together with either (a) a summary description of procedures for sample collection, sample transfer or transport, and sample analysis, or (b) copies of procedures for sample collection, sample transfer or transport, and sample analysis, in accordance with the'proposed review schedule but in no case less than 4 months prior to the issuance of an operating license.

X.1.21.2 Inte retation None required.

X.1.21.3 Statement of Res onse It is intended to comply with this requirement by (1) adding a dedicated post-accident sample station; (2) adding additional instrumentation to the on-site chemistry laboratory; and (3) contracting with an off-site laboratory on a contingency basis for selected chemica'1 and radiochemical analyses. Design documentation will be submitted by fuel load.

Hodifications will be completed by January 1, 1982.

X.1.22 TRAINING FOR MITIGATING CORE DAMAGE (II.B.4)

X.l.22.1 Statement of Re uirement Licensees are required to develop and implement a training program to teach the use of installed equipment and systems to control or mitigate accidents in which the core is severely damaged.

Shift technical advisors and operating personnel from the plant manager through the operations chain to the licensed operators shall receive all the training'ndicated in Table X.l.22-1.

Applicants for operating licenses should develop a training program prior to fuel loading and complete personnel training prior to full-power operation.

X.l.22.2 Inte retation None required.

X.l-30

X.l.22.3 Statement of Res onse The training program has been developed as required above. Training of personnel will be completed prior to a full power license. Appropriate documentation will be made available for review by Region I ISZ.

X.l-31

TABLE X.l.22-1 TRAINING CRITERIA FOR MITIGATING CORE DAMAGE A program is to be developed to insure 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 should include the following topics.

A. Incore Instrumentation

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

2. Methods for calling up (printing) incore data from the plant computer.

B. Vital Instrumentation Instrumentation response in an accident environment; failure sequence (time to failure, method of failure); indication reliability (actual vs. indicated level).

2. Alternative methods of measuring flows, pressures, levels, and temperatures.

a. Determination of reactor pressure vessel level if all level transmitters fail.

b. Determination of other reactor coolant system parameters if the primary method of measurement has failed.

C. Primar Chemistr

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 breakdown for core damage; for clad damage.

3. Corrosion effects of extended immersion .in primary water; time to failure.

D. Radiation Monitorin

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 differenct locations; use of detectors to determine extent of core damage..

X.1-32

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

Gas 'Generation

l. Methods of H generation during an accident; other sources'of gas (Xe, Ke); techniques for venting or disposal of non-condensibles.

2. H> flammability and explosive limit; sources of 0> in containment or reactor coolant system.

\

X.1-33

K.l.23 RELIEF AND SAFETY VALVE TEST REQUIREMENTS (II.D.1)

K.1.23.1 Statement of Re uirement Pressurized-water reactor and boiling-water reactor licensees and applicants shall conduct testing to qualify the reactor coolant system relief and safety valves under expected operating conditions for design-basis transients and accidents. ~

Licensees and applicants shall- determine the expected valve operati'ng conditions through the use of analyses of accidents and anti'cipated operational occurrences referenced in Regulatory Guide 1.70, Revision 2.

The single failures applied to these analyses shall be chosen so that the dynamic forces on the safety and relief valves are maximized. Test pressures shall be the highest predicted by conventional safety analysis procedures: Reactor coolant system relief and safety valve qualificaiton shall include qualification of associated control circuitry, piping', and supports, as well as the valves themselves.

Preimplementation review will be based on EPRI, BWR, and applicant submittals with regard to the various te'st programs. These submittals should be made on a timely basis as noted below, to allow for adequate review and to ensure that the following valve qualification dates can be met:

Final BWR Test Program--October 1, 1980 Block'Valve Qualification Program--January 1, 1981 Postimplementation review will be based on the applicants'lant-specific submittals for qualification of safety relief valves. To properly evaluate these plant-specific applications, the test data and results of the vaiious programs will also be required by the following dates:

BWR Generic Test Program Resutls--July 1, 1981 Plant-specific submittals confirming adequacy of safety and relief valves based on licensee/applicant preliminary review of generic test program results--July 1, 1981 Plant-specific reports for safety and relief valve qualification--

October 1, 1981 Plant-specific submittals for piping and support evaluations--

January 1, 1982 X.l.23.2 Inter retation None required.

X:1.23.3 Statement of Res onse PPSL is participating in the BWR Owner's Group (BWROG) program to test safety/relief valves (SRVs). Wyle Laboratories in Huntsvill'e, Alabama has.

X.1-34

been contracted to design and build a test facility. The design is complete and construction is well underway. The facility will be capable of high and low pressure valve tests.

Documentation of the BWROG testing program was sent to the NRC on September 17, 1980 by, a letter from D.B. Waters 'to,-R.N. Vollmer. A summary of'this document is provided below.

An engineering evaluation was done to identify the expected operating conditions for SRVs during design basis,-transients and accidents. This evaluation indicates the SRVs may be required to pass low pressure liquid as a result of the Alternate Shutdown Mode (described in Subsection 15.2.9).

No other significantly probable event, even combined with a single active failure or single operator error, produces expected operating conditions that justify qualification of SRVs for extreme operating conditions.

Therefore a test program was developed to demonstrate the as may be necessary during the Alternate Shutdown Mode.

SRVs'apabilities PPRL is reviewing the program description and scope. The testing is scheduled for completion by July 1, 1981. A plant specific SRV qualification report will be submitted to the NRC by October- 1981. A plant specific evaluation will be submitted by January 1982.

X.l.24 SAFETY/RELIEF VALVE POSITION INDICATION (II.D.3)

X.1.24.1 Statement of Re uirement Reactor coolant system relief and safety valves shall be provided 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.

Documentation should be provided that discusses each item of the clarification, as well as electrical schematics and proposed te'st procedures in accordance with the proposed review schedule, but in no case less than 4 months prior to the scheduled issuance of the staff safety evaluation report. Implementation must be'completed prior to fuel load.

X.1.24.2 Inter retation None required.

X.l-35

X.1.24.3 Statement of Re uirement Each of the 16 safety/relief valves will be provided with an acoustic monitor to detect flow through the valve. The sensors will be located in the piping downstream of each valve. The monitors will be grouped into two divisions. Each division will have a group annunciator. Individual indication of an open valve will be provided by 16 lights in the control room on a front row panel. This information will also be available on the back row panel where the signal conditioning instruments will be located.

All equipment will be installs'd by fuel load.

K.1.25 AUXILIARYFEEDWATER SYSTEM EVALUATION (II.E.l.l)

This requirement is not applicable to SSES.

X.l.26 AUXILIARYFEEDWATER SYSTEM INITIATION AND FLOW (II.E.1.2)

This requirement is not applicable to SSES.

X.l,27 EtiERGENCY POWER FOR PRESSURIZER HEATERS (II.E.3.1)

This requirement is not applicable to SSES.

X.l.28 DEDICATED HYDROGEN PENETRATIONS (II.E.4.1)

X.1.28.1 Statement of Re uirment Plants using external recombiners or purge systems for postaccident combustible gas control of the containment atmosphere should provide containment penetration systems for external recombiner or purge systems that are dedicated to that service only. These systems must meet the redundance and single-failure requirements of General Design Criteria 54 and 56 of Appendix A to 10 CFR 50, and that are sized to satisfy the flow requirements of the recombiner or purge system.

The procedures for the use of combustible gas control systems following an accident that results in a degraded core and'elease of radioactivity to the containment must be reviewed an revised, if. necessary.

Operating license'applicants must have design 'changes completed by July 1, 1981 or prior to issuance of an operating license, whichever is later.

X.l.28.2 Inter retation None required.

X.1-36

X.l.28.3 Statement of Res onse SSES design includes 100/ redundant internal hydrogen recombiner systems for postaccident combustible gas (hydrogen) control. Therefore this requirement is not applicable to SSES.

X.l.29 CONTAINMENT ISOLATION DEPENDABILITY (II.E.4.2)

X.1.29.1 Statement of Re uirement (1) Containment isolation system designs shall comply with the recommendations of Standard Review Pl'an Section 6.2.4 (i.e., that there be diversity in the parameters sensed for the initiation of containment isolation).

(2) All plant personnel shall given careful consideration to the definition of essential and nonessential systems, identify each system determined to be essential, identify each system determined to be nonessential, describe the basis for selection of each essential system, modify their containment isolation designs accordingly, and report the results of the reevaluation to the NRC.

(3) All nonessential systems shall be automatically isolated by the containment isolation signal.

(4) The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves.

Reopening of containment isolation valves shall require deliberate operator action.

(5) The containment setpoint pressure that initiates containment isolation for nonessential penetrations must be reduced to the minimum compatible with normal operating conditions.

(6) Containment purge valves that do not satisfy the operability criteria set forth in Branch Technical Position CSB 6-4 or the Staff Interim Position of October 23, 1979 must be sealed closed as defined in SRP 6.2.4, item II.3.f during operational conditions 1, 2, 3, and 4.

Furthermore, these valves must be verified to be closed at least every 31 days.

(7) Containment purge and vent isolation valves must close on a high radiation signal.

Applicants for an operating license must be in compliance with positions 1 through 4 before receiving an operating license. Applicants must be in compliance with positions 5 and 7 by July 1, 1981, and position 6 by X.l-37

January 1, 1981 or before they receive their operating license, whichever is later for each position.

X.l.29.2 Inter retations From item 4, the opening of containment isolation valves must require a deliberate operator action.

From item 5, the containment isolation setpoint pressure should be optimized to prevent unnecessary isolations during normal operations.

However, containment isolation must not be prevented or delayed during 'an accident.

From item 7, radioactive materials must not be released from the containment to the environment following an accident.

X.l.29.3 Statement of Res onse (1) Containment isolation is actuated by several sensed parameters (refer to Table 3.3.2-1 in the Technical Specifications). This complies with the Standard Review Plan, Section 6.2 ', Paragraph II-6.

(2) A preliminary evaluation of essential and non-essential systems has been completed. This evaluation is currently being reviewed. A final evaluation will be prepared and submitted to the NRC by liarch 1981.

(3) liodifications will be made (following completion of'the evaluation for item 2 above) to automatically isolate all non-essential systems or justification will be provided. These modifications and/or justifications will be completed prior to fuel load.

(4) All but several NSSS vendor supplied containment isolation valves do not automatically open when a containment isolation signal is reset.

These valises are designed to automatically open.

(5) The containment pressure setpoint is currently being evaluated.

Justification will be provided by Harch 1981.

(6) Branch Technical Position CSB 6-4 was considered in the design of containment purge valves. All valves that do not meet these operability criteria will be verified to be locked closed every 31 days. An administrative procedure will be. written by fuel load in compliance to this requirement.

(7) Containment vent and purge lines are part of the standby gas treatment system (SGTS). The exhaust lines of this system contain radiation monitors. During a high radiation condition, these monitors will produce a signal to trip system fans and close dampers in the exhaust line to isolate the system. This design meets the intent of the requirement.

X.l-38

X.l.30 ACCIDENT-MONITORING INSTREiENTATION (II.F.l)

X.l.30.1 Statement of Re uirement The following equipment shall be added:

(1) Noble gas effluent radiological monitor; (2) Provisions for continuous sampling of plant effluents for postaccident releases of radioactive iodines and particulates and onsite laboratory capabilities; (3) Containment high-range radiation monitor; (4) Containment pressure monitor; (5) Containment water level monitor; and (6) Containment hydrogen concentration monitor.

It is important that the displays and controls added to the control room as a result of this requirement not increase the potential for operator error.

A human-factors analysis should be performed which considers:

(a) the use of this information by an operator during both normal and abnormal plant conditions, (b) integration into emergency procedures, (c) integration into. operator training, and (d) other alarms during emergency and need for prioritization of alarms.

Each piece of equipment is further discussed below.

X.1.30.1.1 Noble Gas Effluent Honitor Noble gas effluent monitors shall be installed with an extended range designed to function during accident conditions as well as during normal operating conditions. Hultiple monitors are considered necessary to cover the ranges of interest.

(1) Noble gas effluent monitors with an upper range capacity of 10 u Ci/cc (Ke-133) are considered to be practical and should be installed in all operating plants.

(2) Noble gas effluent monitoring shall be provided for the total range of concentration extending from normal condition (as low as reasonably achievable concentrations to a maximum of K.l-39

10~@ Ci/cc (Xe-133). Multiple monitors are considered to be necessary to cover the ranges of interest. The range capacity of individual monitors should overlap by a factor of ten.

Licensees and licensing applicants should have available for review the final design description of the as-built system, including piping .and instrument diagrams'together with either (1) a description of procedures for system operation and calibration, or (2) copies of procedures for system operation and calibration. License applicants will submit the above details in accordance with the proposed review schedule, but in no case less than 4 months prior to the issuance of an operating license.

X.1.30.1.2 Sam lin and Anal sis of Plant Effluents Because iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radioiodines for the accident condition shall be provided with sampling conducted by adsorption on charcoal or other media, followed by onsite laboratory analysis.

Licensees shall provide continuous sampl'ing of plant gaseous effluent for postaccident releases of radioactive iodines and particulates to meet. the requirements of Table II.F.1-2 in NUREG 0737. Licensees shall also provide onsite laboratory capabilities to analyze or measure these samples. This requirement should not be construed to prohibit design and development of radioiodine and particulate monitors to provide online sampling and analysis for the accident condition. If gross- gamma radiation measurement techniques are used, then provisions shall be made to minimize noble gas interference.

The shielding design basis is given. in Table II.F.1-2 of HUREG 0737. The sampling system design shall be such that plant personnel could remove samples, replace sampling media and transport the samples to the onsite analysis facility with radiation exposures that are not in excess of the criteria of GDC 19 of 5-rem whole-body exposure and 75 rem to the extremities during the duration of the accident.

The design of the systems for the sampling of particulates and iodines should provide for sample nozzle entry velocities which are approximately isokinetic (same velocity) with expected induct or instack air velocities.

For accident conditions, sampling may be complicated by a reduction in stack or vent effluent velocities to below design levels, making it necessary to substantially reduce sampler intake flow rates to achieve the isokinetic condition. Reductions in air flow may well be beyond the capability of available sampler flow controllers to maintain isokinetic conditions; therefore, the staff will accept flow control devices which have the capability of maintaining isokinetic conditions with variations in stack or duct design flow velocity of +20%. Further depart'ure from the isokinetic condition need not be considered in design. Corrections for

'I X. 1-40

non-isokinetic sampling conditions, as provided in Appendix C of ANSI 13.1-1969 may be considered on an ad hoc basis.

Effluent streams which may contain air with entrained water, e.g. air ejector discharge, shall have provisions, e.g., heaters, to ensure that the adsorber is not degraded while providing a representative sample.

License applicants will submit final design details in accordance with the proposed review schedule, but in no case less than 4 months prior to the issuance of an operating license.

X.l.30.1.3 Containment Hi h-Ran e Radiation Monitor In containment radiation-level monitors with a maximum range of 10 8 rad/hr shall be installed. A minimum of two such monitors that are physically separated. shall be provided. Monitors shall be. developed and qualified to function in an accident environment.

The specification of 10 rad/hr in the above position was based on a calculation of postaccident containment radiation levels that include both particulate (beta) and photon (gamma) radiation. A radiation detector that responds to both beta and gamma radiation cannot be qualified to post-IOCA (loss-of-coolant accident) containment env'ironments but gamma-sensitive instruments can be so qualified. In order to follow the course of an accident. A containment monitor that measures only gamma radiation is adequate. The requirement was revised in the October 30, 1979 letter to provide for a photon-only measurement with an upper range of 10 R/hr.

The monitors shall be located in containment(s) in. a manner as,to provide a reasonable assessment of area radiation conditions inside containment. The monitors shall be widely separated so as to provide independent measurements and shall "view" a large fraction of the containment volume.

Monitors should not be placed in areas which are protected by massive shielding and should be reasonably accessible for replacement, ma'intenance, or calibration. Placement high in a reactor building dome is not recommended because of potential maintenance difficulties.

The monitors are required to respond to gamma photons with energies as low as 60 keV and to provide an essentially flat response for gamma energies between 100 keV and 3 MeV, as specified in Table II.P.1-3 of NUREG 0737.

Monitors that use thick shielding to increase the upper range will under-estimate postaccident radiation levels in containment by several orders of magnitude because of their insensitivity to low energy gammas and are not acceptable.

License applicants will submit the required documentation in accordance with the appropriate review schedule, but in no case less than 4 months prior to the issuance of the staff evaluation report for an operating license.

X.l.30.1.4 Containment Pressure Monitor A continuous indication of containment pressure shall be provided in the control room of each operating reactor. Measurement and indication capability shall include three times the design pressure of the containment for concrete, four times the design pressure for steel, and -5 psig for all containments.

Operating license applicants with an operating license dated before January 1, 1982 must have design changes completed by January 1, 1982; those applicants with license dated after January 1, 1982 must,hav'e all design modifications completed before they can receive their operating license.

Documentation is due 6 months for the expected date of operation..

X.l.30.1.5 Containment Water Level Monitor A continuous indication of containment water level shall be provided in the control room for all plants. A wide range instrument shall be provided to cover the range from the bottom to 5 feet above the normal water level in the suppression pool.

The containment wide-range water level indication channels shall meet appropriate design and qualification criteria. The narrow-range channel shall meet the requirements of Regulatory Guide 1.89.

For BWR pressure-suppression containments, the emergency core cooling system suction line inlets may be used as a starting reference point for the narrow-range and wide-range water level monitors, instead of the bottom of the suppression pool.

The accuracy requirements of the water level monitors shall be provided and justified to be adequate for their intended function.

Operating license applicants with an operating license date before July 1, 1981 must have design changes completed by July 1, 1981, whereas those applicants with license dates past July 1, 1981 must have all design modifications completed before they can receive their operaitng license.

Submittals from operating reactors licensees and applicants for operating licenses (with an operating license date before January 1, 1982) shall be pr'ovided by January 1, 1982. Applicants with operating license dates beyond January 1, 1982 shall provide the required design information at least 6 months before the expected date of operation.

X.1.30.1.6 Containment H dro en Monitor A continuous indication of hydrogen concentration in the containment atmosphere shall be provided in the control room. Measurement capability shall be provided over. the range of 0 to 10% hydrogen concentration under both positive and negative ambient pressure.

Operating license applicants with an operating license date before January 1, 1982 must have design changes completed by January 1, 1982 must have all design modifications completed before they. c'n receive their operating license.

Operating reactors and applicants for'operating license receiving license before January 1, 1982 will submit documentation before an'perating January 1, 1982. Applicants with operating license issued after January 1, 1982 shall provide the required design information at least 6 months prior to the expected date of operation.

X.l.30.2 Inter retation None required.

X.l.30.3 Statement of Resonse The response for 'each equipment requirement is given below. All equipment will be installed by the required dates. A human factors evaluation will be performed for changes that involve control room instrumentation.

K.l.30.3.1 Noble Gas Effluent Monitor Two overlapping high range channels will be, added to each plant vent channels will monitor isokinetic samples taken from each vent. Each path.'hese vent will have the capability of monitoring 'Xe-133 up to a concentration of 10 p Ci/cc. This information will be available in the control room and the Technical support center.

X.l.30.3.2 Sam lin and Anal sis of Plant Effluents A continuous isokinetic sample is taken from each plant effluent vent path.

The sample stream is run through a HEPA filter and then through a charcoal filter. Scintilation detectors view both filters. The output from the detector viewing the charcoal filter is passed through a single channel analyzer which is adjusted to detect I-131. Both filters can be quickly removed for on-site or mobile analysis.

X.l.30.3.3 Containment Hi h-Ran e Radiation Monitor Two redundant high range radiation monitors will be installed in the primary containment. These detectors were designed to meet the requirements of NUREG-0578. The output of each monitor is recorded in the control room. The monitors are capable of measuring up to 1 x 10 " R/hr (gamma only).

X.l.30.3.4 Containment Pressure Monitor Containment pressure will be monitored by two sets of redundant instruments. One set has a range of 0-65 PSIA and the other has a range of X.l-43

0-250 PSIG. These ranges cover the requir'ed range of 3 time containment design pressure to -5 psig. The outputs of these instruments are recorded in the control room.

X.l.30.3.5 Containment Water I,evel Monitor Two redundant wide range level instruments monitor the level of water in the suppression pool. The range of the instruments extends from 4.5 feet to 49 feet above the bottom of the suppression pool (normal level is 23 feet). The output of these instruments are recorded in the control room.

X.l.30.3.6 Containment H dro en Monitor Two redundant hydrogen instruments monitor the containment atmosphere.

These instruments cover a range of 0 to 10/ hydrogen. The output of each instrument'is recorded in the control room.

X.l.31 INSTRUMENTATION FOR DETECTION OF INADE UATE CORE COOLING (II.F.2)

X.1.31.1 Statement of Re uirement Iicensees shall provide a description of any additional instrumentation or controls (primary or backup) proposed for the plant to supplement existing instrumentation (including primary coolant saturation'onitors) in order to provide an unambiguous, easy-to-interpret indication of inadequate core cooling (ICC). A description of the functional design requirements for the system shall also be included. A description-of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided.

X.l.31.2 Inter retation None required.

X.l.31.3 Statement .of Res onse As stated in NEDE-24801, "Review of BWR Reactor Vessel Water Level Measurement", the optimum design is the cold reference leg configuration with parallel instrument lines. Since the present design conforms to this recommended configuration, no additional instrumentation or modifications to instrumentation are needed for detection of inadequate core cooling (ICC).

Procedures are being developed (in response to requirement I.C.l) for proper identification of ICC. Refer to subsection X.l.8 for the response to requirement I.C.l.

PPRL has decided to provide an additional computer display format which will promote detection of ICC with the existing instrumentation. A discussion of this display format will be provided by fuel load, X.l-44

X.l.32 EMERGENCY POWER FOR PRESSURIZER EQUIPMENT (II.G.1)

This requirement is not applicable to SSES.

X.l.33 REVIEW ESF VALVES (II.K.1..5)

No requirement stated in NUREG 0737.- Refer to Subsection X.2.25 which contains the response to the requirement in NUREG 0694.

X.1.34 OPERABILITY STATUS (II.K.1.10)

No requirement stated in NUREG 0737. Refer to Subsection X.2.26 which contains the response to the requirement in NUREG 0694.

X.l.35 TRIP PRESSURIZER LOW-IEVEL COINCIDENT SIGNAL BISTABLES (II.K.1.17)

This requirement is not applicable to SSES.

X.l.36 OPERATOR TRAINING FOR PROMPT MANUAL REACTOR TRIP (II.K.1.20)

This requirement is not applicable to SSES.

X.l.37 AUTOMATIC SAFETY GRADE ANTICIPATORY REACTOR TRIP (II.K.1.21)

This requirement is not applicable to SSES.

X.1.38 AUXILIARYHEAT REMOVAL SYSTEM PROCEDURES (II.K.1.22)

No requirement stated in NUREG 0737.'efer to Subsection X.2.30 which contains the response to the requirement in NUREG 0694.

X.l.39 REACTOR VESSEL LEVEL PROCEDURES (II.K.1.23)

No requirement stated in NUREG 0737. Refer to Subsection X.2.31 which contains the response to the requirement in NUREG 0694.

X.l.40 COMMISSION ORDERS ON BABCOCK AND WILCOX PLANTS (II.K.2)

These requirements are not applicable to SSES.

X.l-45

X.l.41 AUTOMATIC POWER-OPERATED RELIEF'VALVE ISOIATION SYSTEM (II.K.3.1)

This requirement is not applicable to SSES.

X.1.42 REPORT ON POWER-OPERATED RELIEF VALVE FAILURES (II.K.3.2)

This requirement is not applicable to SSES.

X.1.43 REPORTING SAFETY/RELIEF VALVE FAILURES AND CHALLENGES (II.K.3.3)

No requirement stated in NUREG 0737. Refer to Subsection X.2.33. which contains the response to the requirement in NUREG 0694.

X.1.44 AUTOMATIC TRIP OF REACTOR COOIANT PUMPS DURING A LOCA (II'.K.3.5)

This requirement is not applicable to SSES.

X.1.45 EVALUATION OF POWER-OPERATED RELIEF VALVE OPENING PROBABILITY II.K.3.7 This requirement is not applicable to SSES.

X.l.46 PROPORTIONAL INTEGRAL DERIVATIVE CONTROLLER MODIFICATION (II.K.3.9)

This requirement is not applicable to SSES.

X.1.47 PROPOSED ANTICIPATORY TRIP MODIFICATION (II.K.3.10)

This requirement is not applicable to SSES.

X.1.48 POWER-OPERATED RELIEF VALVE FAILURE RATE (II.K.3.11)

This requirement is not applicable to SSES.

X.l.49 ANTICIPATORY REACTOR TRIP ON TURBINE TRIP (II.K.3.12)

This requirement is not applicable.to SSES.

X.l-46

X.l.50 SEPARATION OF HIGH PRESSURE COOLANT INJECTION AND REACTOR CORE ISOLATION COOLING SYSTEM INITIATION LEVELS II.K.3.13 X.1.50.1 Statement of Re uirement Currently, .the reactor core isolation 'cooling (RCIC) system and the high-pressure coolant injection (HPCI) system both initiate on the same low-water-level signal and both isolate. on the same high-water-level signal.

The HPCI system will restart on low water level but the RCIC system will not. The RCIC system is a low=flow system when compared'to the HPCI system. .The initiation levels of the HPCI arid RCIC system should be separated so that the RCIC system initiates at a higher water level than the HPCI system. Further, the initiation 'logic of the RCIC system should be modified so that the RCIC system will restart on low water level. These changes have the potential to reduce the number of challenges to the HPCI system and could result in less stress on the vessel from cold water injection. Analyses should be performed to evaluate these changes. The analyses should be submitted to the NRC staff and changes should be implemented if justified by the analyses.

All applicants for operating license should submit the results of an evaluation and proposed modifications 4 months prior to the expected issuance of the staff safety evaluation report for an operating license or 4 months prior to the listed implementation date, whichever is later.

X.1.50.2 Inter retation None required.

X.1.50.3 Statement of Res onse The BWR Owners'roup (BWROG) has performed a generic evaluation in response to this requirement. They have stated the automatic reset of the RCIC system following a high water level trip will improve the overall safety of the BWR. PPSL is currently reviewing the BWROG evaluation and will prepare a response upon completion.

X.l.51 MODIFY BREAK-DETECTION LOGIC TO PREVENT SPURIOUS ISOLATION OF HIGH PRRESSURE COOLANT INJECTION AND REACTOR CORE ISOLATION COOLING II.K.3.15 X.l.51.1 Statement of Re uirement The high-pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) systems use differential pressure sensors on elbow taps in the steam lines to their turbine drives to detect and isolate pipe breaks in the systems. The pipe-break-detection circuitry has resulted in spurious 'isolation of, the HPCI and RCIC systems due to the pressure spike which accompanies startup of the systems'. The pipe-break-detection X.1-47

circuitry should be modified so that pressure spikes resulting from HPCI and RCIC system initiation will not cause inadvertent system isolation.

All applicants for operating license should submit documentation 4 months prior to the expected issuance of the staff safety evaluation report for an operating license or 4 months prior to the listed implementation date, whichever is later.

Inter retation

'.l.51.2 None required.

X.l.51.3 Statement of Res onse The BWR Owners'roup has evaluated possible modifications and recommends the installation of timing devices in the pipe-break-circuitry. These devices will prevent inadvertent isolations following short periods'f high flow rates which occur from a pressure spike. A sustained high flow rate, which will occur following a pipe break, will cause a proper isolation to occur.

PPGL is pursuing this recommendation. It is anticipated that design-information will be submitted by April 1981 and equipment will be installed by fuel load, pending availability.

X.1.52 REDUCTION OF CHALLENGES AND FAILURES OF RELIEF 'VALVES (II.K.3.16)

X.l.52.1 Statement of Re uirement The record of relief-valve failures to close for all boiling-water reactors (BWRs) in the past 3 years of plant operation is approximately 30 in 73 reactor-years (0.41 failures per reactor-year). This has demonstrated that the failure of a relief valve to close would be the most likely .cause of a small-break loss-of-coolant accident (IOCA). The high failure rate is the result of a high relief-valve challenge rate and a relatively high failure rate per challenge (0.16 failures per challenge). Typically, five valves are challenged in each event. This results in an equivalent failure rate per challenge of 0.03. The challenge and failure rates can be reduced in the following ways:

(1) Additional anticipatory scram on loss of feedwater, (2) Revised relief-valve actuation setpoints, (3) Increased emergency core cooling (ECC) flow, (4) Lower operating pressures, (5) Earlier initiation of ECC systems X.1-48

(6) Heat removal through emergency condensers, (7)'ffset valve setpoints to open fewer ya'ives per challenge, (8) Installation of additional relief valves with a block- or isolation-valve feature to eliminate opening, of the safety/relief valves (SRVs),

consistent with the ASME Code, (9) Increasing the high steam line flow setpoint for main steam line isolation valve (MSIV) closure, (10) I,owering the pressure setpoint for MSIV closure, (ll) Reducing the testing frequency of the MSIVs, (12) More-stringent valve leakage criteria, and (13) Early removal of leaking valves An investigation of the feasibility and contraindications of reducing challenges to the relief valves by use of the aforementioned methods should be conducted. Other methods should also be included in the feasibility study. Those changes which are shown to reduce relief-valve challenges without compromising the performance of the relief valves or, other systems should be implemented. Challenges to the relief valves should be reduced substantially (by an order of magnitude).

Results of the evaluation shall be submitted by April 1, 1981 for staff review. The actual modificaiton shall be accomplished during the next scheduled refueling outage. following staff approval or no later than 1 year following staff approval. Modification to be implemented should be documented at the time of implementation.

X.l.52.2 Inter retation None required.

X.1.52.3 Statement of Res onse The BWR Owners'roup (BMROG) is developing recommendations to comply with this requirement. These recommendations are scheduled to be available for review by March 1981. PP8<L will prepare a final response following review of the BWROG report.

X.1.53 REPORT ON OUTAGES OF EMERGENCY CORE COOLING SYSTEMS (II.K.3.17)

X.1.53.1 Statement of Re uirement Several components of the emergency core-cooling (ECC) systems are permitted by technical specifications to'ave substantial outage times X.l-49

(e.g., 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for one diesel-generator; 14 days for the HPCI system). In addition, there are no cumulative outage time limitations for ECC systems.

Licensees should submit a report detailing outage dates and lengths of outages for all ECC systems for the last 5 years of operation. The report should also include the causes of the outages (i.e , controller failure, spurious isolation),

X.l.53.2 Inter retation Licensees should provide a report which contains emergency core cooling system unavailability data. This requirement can not be, applicable to SSES, since the plant has never operated.

X.1.53.2 Statement of Res onse This requirement is not applicable to SSES.

X.l.54 MODIFICATION OF AUTOMATIC DEPRESSURIZATION SYSTEM LOGIC (II.K.3.18)

X.1.54.1 Stat'ement of Re uirement The automatic depressurization system (ADS) actuation logic should be modified to eliminate the need for manual actuation to assure adequate core cooling. A feasibility and risk assessment study is required to determine the optimum approach. One possible scheme that should be considered is ADS actuation'n low reactor-vessel water level provided no high-pressure coolant injection or high-pressure coolant system flow exists and a low-pressure emergency core cooling system is running. This logic would complement, not replace, the existing ADS actuation logic.

Applicants for operating license shall provide results of feasibility study 1 year prior to issuance of operating license. A description of the proposed modification for staff approval is required 4 months prior to issuance of an operating license.

X.1.54.2 Inter retation The ADS actuation logic will not be actuated for steam line breaks (SLB) outside containment. The operator must manually actuate the ADS after diagnosing that an SLB has occurred. The ADS actuation logic should be modified to provide automatic actuation for all Design Basis Accidents.

X.l,54.3 Statement of Res onse The BVR Owners'roup i's currently performing a generic feasibility study.

The results of this effort will be reviewed and implemented,as appropriate for SSES.

X.1-50

X.l.55 RESTART OF CORE SPRAY AND LOW PRESSURE COOLANT INJECTION SYSTEMS II.K.3.21 X.l.55.1 Statement of Re uirement The core-spray and low-pressure, coolant;injection (LPCI) system flow may be stopped by the operator. These systems will not restart automatically on loss of water level if an initiation signal is still present. The core spray and LPCI system logic should be modified so that these systems will restart, if required, to assume adequate core cooling. Because this design modification affects several" core-cooling modes under accident conditions, a preliminary design should be submitted for staff review and approval prior to making the actual modification.

All applicants for operating license should submit documentation 4 months prior to the expected issuance of an operating license or 4 months prior to the listed implementation date, whichever is later.

X.1.55.2 Inter retation None required.

X.l.55.3 Statement of Res onse The BWR Owner's Group (BWROG) has performed a generic evaluation in response to this requirement. PPGL will prepare a response following review of the BWROG report.

X.l.56 AUTOMATIC SWITCHOVER OF REACTOR CORE ISOIATION COOLING SYSTEM SUCTION II.K.3.22'.l.56.1 Statement of Re uirement The reactor core isolation cooling (RCIC) system takes suction from the condensate storage tank with manual switchover to the suppression pool when the condensate storage tank level is low. This switchover should be made automatically. Until the automatic switchover is implemented, licensees should verify that clear and cogent procedures exist for the manual switchover of the RCIC system suction from'the condensate storage tank to the suppression pool.

S Documentation must be submitted 4 months prior to issuance of the staff safety evaluation report or 4 months prior to the implementation date, whichever is later. Modifications shall be completed by January 1, 1982.

X.1.56.2 Inter retation None required.

X.l.56.3 Statement of Res onse Procedures outlining the manual switchover of RCIC suction on condensate storage tank low level to the suppression pool will be available by fuel load. Appropriate documentation will be made available for review by Region I ISE. The design changes for automatic switchover are being developed. All modifications will be completed by January 1982.

X.l.57 CONFIRM ADE UACY OF %PACE COOLING FOR HIGH PRESSURE COOLANT INJECTION AND REACTOR CORE ISOLATION COOLING SYSTEMS II.K.3.'24)

X.1 57.1

~ Statement of Re uirement Long-term operation of the reactor core isolation .cooling (RCIC) and high-pressure coolant injection (HPCI) systems may require space cooling to maintain the pump-room temperatures within allowable limits. Licensees should verify the acceptability 'of the consequences of a complete loss of alternating-current (AC) power. The RCIC and HPCI systems should'e designed to withstand a complete loss of offsite AC power to their support systems, inclu'ding coolers, for at least' hours.

All applicants for operating license should submit documentation 4 months prior to the expected issuance of the staff safety evaluation report foi an operating license or 4 months prior to the listed implementation date, whichever is later.

X.1.57.2 Inter retation Confirm that HPCI and RCIC room cooling can be maintained to enable continuous operation during a loss of offsite AC power for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

X.1.57.3 Statement of Res onse The HPCI and RCIC room unit coolers are designed to withstand the consequences of a complete loss of offsite AC power since these are powered from onsite diesel generators. In addition, 100 percent redundant unit coolers are provided in both of these systems we believe this to be true with all the other HPCI and RCIC support systems. This will be confirmed prior to fuel load.

X.l.58 EFFECT OF LOSS OF ALTERNATING-CURRENT POWER ON RECIRCULATION PUMP SEAIS II.K.3.25 X.l.58.1 Staement of Re uirement The licensees should determine, on a plant-specific basis, by analysis or experiment, the consequences of a loss of cooling water to the reactor recirculation pump seal coolers. The pump seals should be designed.to X.1-52

withstand a complete loss of alternating-current (AC) power for at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Adequacy of the seal design should be demonstrated.

Applicants for operating licenses shall submit the evaluation and proposed modifications no later than 6 months prior to expected issuance of the staff safety evaluation report in support of license issuance, whichever is later. Modifications must be completed by January 1, 1982.

X.1.58.2 Inte retation Evaluate the effect of a loss of offsite AC gower for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months on the recirculation pump seals.

X.1.58.3 Statement of Res onse The BWR Owners'roup (BWROG) is performing an evaluation in response to this requirement. PPGL will prepare a response following review of the BVROG report, which is scheduled for completion by April 1, 1981.

Preliminary results indicate a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months loss of offsite AC power does not produce a significant impact to the safe operation of the plant.

X.l.59 PROVIDE A COMMON REFERENCE LEVEL'OR VESSEL LEVEL INSTRlkiENTATION II.K.3.27 X.1.59.1 Statement of Re uirement Different reference points of the various reactor vessel water level instruments may cause operator confusion. Therefore, all level instruments should be referenced to the same point. Either the bottom of the vessel or the top of the active fuel are reasonable reference points.

All applicants for operating license should submit documentation 4 months prior to the expected issuance of the staff safety evaluation report for an operating license or 4 months prior to the listed implementation date, whichever is later.

X.1.59.2 Inter retation None required.

X.l.59.3 Statement of Res onse The BVR Owners'roup (BMROG) has performed a generic evaluation for this requirement. They have stated no changes are necessary and the present system is fully adequate to allow plant operators to respond properly under all postulated reactor conditions. PPGL has reviewed the BWROG evaluation and concurs with its findings. Therefore no changes will be made to vessel level instrumentation..

X.1-53

X.l.60 VERIFY QUALIFICATION OF ACCUMULATORS ON AUTOHATIC DEPRESSURIZATION SYSTEM VALVES II.K.3.28 X.l.60 ' Statement of Re uirement Safety analysis reports. claim that air or nitrogen accumulators for the automatic depressurization system (ADS) valves are provided with sufficient capacity to cycle the valves open five times at design pressures. GE has also stated that the emergency core cooling (ECC) systems are designed to withstand a hostile environmertt and still perform their function for 100 days following an accident. Licensee should verify that. the accumulators on the ADS valves meet these requirements, even considering normal leakage.

If this cannot be demonstrated, the licensee must show that the accumulator design is still acceptable.

The ADS valves, accumulators, and associated equipment and instrumentation must be capable of performing their functions during and following exposure to hostile environments and taking no credit for nonsafety-related equipment or instrumentation. Additionally', air (or nitrogen) leakage through valves must be accounted for in order to assure that enough inventory of c'ompressed air is available to cycle the ADS valves.

All applicants for operating license shall submit documentation 4 months before the expected issuance of the staff safety evaluation report for an operating license or 4 months before the listed implementation date, whichever is later.

X.1.60.2 Inter retation None required.

X.l.60.3 Statement of Res onse The BWR Owners'roup is performing a generic evaluation in response to this requirement. PPSL is participating in this evaluation, Upon completion, PPSL intends to review the results for applicability and impact to SSES. Implementation of modifications, if needed, will be completed by January 1982.

X.1.61 REVISED SHALL-BREAK LOSS OF COOLANT ACCIDENT METHODS (II.K.3.30)

X.l.61.1 Statement of Re uirement The analysis methods used by nuclear steam supply system vendors and/or fuel suppliers for small-break loss-of-coolant accident (IOCA) analysis for c'ompliance with Appendix K to 10 CFR Part 50 should be'evised, documented and submitted for NRC approval. The revisions should account for comparisons with experimental data, including data from the IOFT test and Semiscale Test facilities.

X.l-54

The Bulletins and Orders Task Force identified a number of concerns regarding the adequacy of certain features of small-break LOCA models, particularly the need to confirm specific model features (e.g.,

condensation heat transfer rates) against applicable experimental data.

These concerns, as they applied to each light-water reactor (LMR) vendor's models, were documented in the task .force also concluded that, in light of the THI-2 accident, additional systems verification of the small-break LOCA model as required by II.4 of Appendix K to 10 CFR 50 was needed. This included providing mental verification of the various modes of single-phase and two-phase natural circulation predicted to occur in each vendor's reactor during small-break LOCAs.

Based on the cumulative staff requirements for additional small-break LOCA model verification, including both integral system and separate effects verification, the staff considered model revision as the appropriate method for reflecting any potential upgrading of the analysis methods.

The purpose of the verification was to provide the necessary assurance that the small-break LOCA models were acceptable to calculate the behavior and consequences of small primary system breaks.'he staff believes that this assurance can alternatively be provided, as appropriate, by additional justification of the acceptability of present small-break LOCA models to speci,fic staff concerns and recent test data. Such justification with'egard could supplement or supersede the need for model revision.

The specific staff concerns regarding small-'break LOCA models are provided in the analysis sections of the BGO Task Force reports for each LMR vendor, (NUREG-0635, -0565, -0626, -0611, and -0623). These concerns should be reviewed in total by each holder of an approved emergency core cooling system model and addressed. in the evaluation as appropriate.

The recent tests include the entire Semiscale small-break test series and LOFT Tests (L3-1) and L3-2). The staff believes that the present small-break LOCA models can be both qualitatively and quantitatively assessed agaist these tests. Other separate effects tests (e.g., ORNL core uncovery tests) and future tests, as appropriate, should also be factored into this assessment.

Based on the preceding information, a detailed outline of the proposed program to address this issue should be submitted. In particular, this submittal should identify (1) which areas of the models, if any, the licensee intends to upgrade, (2) which areas the licensee intends to address by further justification of acceptability, (3) test data to be used as part of the overall verification/upgrade effort, and (4) the estimated schedule for performing the necessary work and submitting this information for staff review and approval.

Licensees shall submit an outline of a program for model justification/revision by November 15, 1980. Licensees shall submit additional information for model justification and/or revised 'analysis

model for staff approval by January 1, 1982. Licensees shall submit their plant-specific analyses using the revised models by January 1, 1983 or one year after any model revisions are approved. Applicants shall submit appropriate information in accordance with the licensing review schedule.

X.l.61.2 Inte retation None required.

X.l.61.3 Statement Of Res onse PPK considers that the reactor vendor, General Electric, is the most appropriate party to work with the staff in resolving staff concerns with small break LOCA models for BWRs. Accordingly, the staff should direct their questions regarding the scope and schedule for this requirement to General Electric (attn. R.'. Buchholz, Manager, BWR Systems Licensing).

Copies of correspondence on this item should be sent to PPSrL so that we may remain cognizant of the progress of the program to resolve the staff's concerns on this requirement. GE has informed us that they have been prep'ared to discuss the approach to this item since December 15, 1980.

X.l.62 PLANT"SPECIFIC CAI,CULATIONS TO SHOW COMPLIANCE WITH 10CFR PART 50.46 II.K.3.31 X.l.62.1 Statement of Re uirement Plant-specific calculations using NRC-approved models for small-break loss-of-coolant accidents (LOCAs) as described in item II.K.3.30 to show compliance with 10 CFR 50.46 should be submitted for NRC approval by all licensees.

X.1.62.2 Inter retation None required.

X.1.62.3 Statement of Res onse Plant specific calculations will be performed following NRC approval of LOCA model revisions required by item II.K.3.30 (see Subsection X.1.61).

X.l.63 EVALUATION OF ANTICIPATED TRANSIENTS WITH SINGLE FAILURE TO VERIFY NO FUEL CLADDING FAILURE (II.K.3.44 X.1.63.1 Statement of Re uirement For anticipated transients combined with the worst single failure an assuming proper operator actions, licensees should demonstrate that the core remains covered or provide analysis to show that no significant fuel, X.l-56

damage results from core uncovery. Transients which result from a stuck-open relief valve should be included in this category.

All applicants for operating license should submit documentation 4 months prior to the expected issuance'of the staff safety evalaution report for an operating license or 4 months prior to the listed implementation date, whichever is later.

X.1.63.2 Inte retation None required.

K.l.63.3 Statement of Res onse The BWR Owners'roup has prepared a generic response to this requirement.

This response contains an evaluation of analyses performed to demonstrate the core remains covered or no significant fuel damage occurs from an anticipated transient with a single failure. PPGL is reviewing this report for applicability to SSES.

X.1.64 EVALUATION OF DEPRESSURIZATION WITH OTHER THAN THE AUTOMATIC DEPRESSURIZATION SYSTEM II.K.3.'45 P

K.l.64.1 Statement of Re uirement Analyses to support depressurization modes other than full actuation of the automatic depressurization system (ADS) (e.g., early blowdown with one or two safety relief valves) should be provided. Slower depressurization would reduce the possibility of exceeding vessel integrity limits by rapid cooldown.

All applicants for operating license should submit documentation 4 months prior to the expected issuance of the staff safety evaluation report for an operatign license or 4 months prior to the listed implementation date, whichever is later.

X.1.64.2 Inter retation None required.

K.l.64.3 Statement of Res onse The BWR Owners'roup (BWROG) has prepared a generic response to this requirement. This response provides the results of an evaluation of depressurization methods. It states that no appreciable improvement can be gained by a slower depressurization and can possibly be detrimental to core cooling. PPSL will prepare a response after reviewing the BWROG report.

X.1-57

K.1.65 MICHELSON CONCERNS (II.K.3.46)

K.l.65.1 Statement of Re uirement A number of concerns related to decay heat removal following a very small break LOCA and other related items were questioned by Mr. C. Michelson of the Tennessee Valley Authority. These concerns were identified for PWRs.

GE was requested to evaluate these concerns as they apply to BWRs and to.

assess the importance of natural circulation during a small-break LOCA in BWRs.

X.l.65.2 Inte retation None required.

X.l.65.3 Statement of Res onse The General Electric Company has responded to the questions posed by Mr.

Michelson. This response was sent by letter from R. H. Buchholz 'to D. F.

Ross'n February 21, 1980. These responses are applicable to SSES and no further response is necessary.

X.1.66 EMERGENCY PREPAREDNESS-SHORT TERM (III.A.1.1)

No requirement stated in NUREG 0737. Refer to Subsection X.2.38 which contains the response to the requirement in NUREG 0694.

X.l.67 UPGRADE EMERGENCY SUPPORT FACILITIES (III.A.1.2)

X.l.67.1 Statement of Re uirement Requirement. to be issued in NUREG 0696.

X.1.67.2 Inter retation None required.

X.l.67.3 Statement of Res onse The response to this requirement will be incorporated into Appendix I of the Emergency Plan; X.1-58

X.1.68 EMERGENCY PREPAREDNESS-LONG TERM (III.A.2)

X.l.68.1 Statement of Re uirement Each nuclear facility shall upgrade its emergency plans to provide reasonable assurance that adequate protective measures can and will be taken in the event of a, radiological emergency.'pecific criteria to meet this requirement is delineated in NUREG-0654 (FEMA-REP-l), "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparation in Support of Nuclear Power Plants."

NUREG-0654, Revision 1; NUREG-0696, "Functional Criteria for Emergency Response Facilities;" and the amendments to 10 CFR Part 50 and Appendix E to 10 CFR Part 50 regarding emergency peparedness, provide more detailed criteria for emergency plans, design, and functional criteria for emergency response facilities and establishes firm dates for submission of upgraded emergency plans for installation of prompt notification systems. These revised criteria and rules supersede previous Commission guidance for the upgrading of emergency preparedness at nuclear power facilities.

Requirements of the new emergency-preparedness rules under paragraphs 50.47 and 50.54 and the revised Appendix E to Part 50 taken together with NUREG-0654 Revision 1 and NUREG-0696, when approved for issuance, go beyond the previous requirements for meteorological programs. To provide a frame for implementation, a staged schedule has been established .with realistic'ime compensating actions provided for interim measures.

Specific milestones have been developed and are presented below.

Milestones are numbered and tagged with the following code; a-date, b-

~activit , c-minimum aces tance criteria. They are as follows:

(1) a. Fuel load.

b. Submittal of radiological emergency respone plans.

a

c. A description of he plan to include elements of HUREG-0654, Revision 1, Appendix 2.

(2) a. Fuel load.

b. Submittal of implementing procedures.

c. Methods, systems, and equipment to assess and monitor actual or potential offsite consequences of a radiological emergency condition shall be provided.

(3) a'. April 1, 1981.

b. Implementation of radiological emergency response plans.

X.l-59

c. Pour elements of Appendix 2 to NUREG-0654 wth the exception of the Class B model of element 3, or Alternative to item (3) requiring compensating actions:

A meteorological measurements program which is consistent with existing technical specifications as the the baseline or an 'he element 1 program and/or element 2 system of Appendix 2 to NUREG-0654, or two independent element 2 systems shall provide the basic meteorologica> parameters (wind direction and speed and an indicator or atmospheric stability) on display,.in the control room. An operable dose calculational methodology (DCM) shall be in use in the control room and at appropriate emergency response facilities.

The following compensating actions shall be taken by the licensee for this alternative:

If only element 1 or element'2 is in use:

o The licensee (the person who will be responsible for making offsite dose projections) shall check communications with the cognizant'ational Weather Service (NWS) first order station'nd NWS forecasting station on a monthly basis to ensure that routine meteorological observations and forecasts can be accessed.

o The licensee shall calibrate the meteorological measurements program at a frequency no less than quarterly and identify a readily available source of meteorological data (characteristic of site conditions) to which they can gain access during calibration periods.

o During conditions of measurements system unavailability, an alternate source of meteorological data which is characteristic of site conditions shall be identified to which the licensee can gain access.

o The licensee shall maintain a site inspection schedule for evaluation of the meteorological measurements program at a frequency no less than weekly.

o It shall be a reportable occurence the if meteorological data unavailability exceeds the goals outline in Proposed Revision 1 to Regulatory Guide 1.23 on a quarterly basis.

X.1-60

(ii) The portion of the DCM relating to the transport and diffusion of gaseous effluents shall be consistent with the characteristics of the Class A model outlined in element 3 of Appendix 2 to NUREG-0654.

(iii) Direct telephone access to the individual responsible for making offsite dose projections '(Appendix E to 10 CFR Part 50(IV)(A)(4)) shall be available to the NRC in the event of a radiological emergency. Procedures for establishing contact and identification of contact individuals shall be provided as part of the implementing procedures.

This alternative shall not be exercised after July 1, 1982. Further, by July 1, 1981, a functional description of the upgraded programs (four elements) and schedule for installation and full operational capability shall be provided (see milestones 4 and 5).

(4) a. March 1, 1982.

b. Installation of Emergency Response Facility hardware and software.

c. .Four elements of Appendix 2'o NUREG-0654, with exception of the Class B model of element 3.

(5) a. July 1, 1982.

b. Full operational capability of milestone 4.

c ~ The Class A. model (designed to be used out to the plume exposure EPZ)'may be used in lieu of Class B model out to the ingestion EPZ. Compensating actions to be taken for extending the application of the Class A model out to the ingestion EPZ include access to supplemental information.

(meso and synoptic scale) to apply judgment regarding intermediate and long-range transport estimates. The distribution of meteorological information by the licensee should be as follows by July 1, 1982:

NRC and Emergency Meteorological Response Organiza-Information CR TSC EOF tions Basic Met. Data X X X (NRC)

(e.g., 1.97 Parameters)

Full Met. Data X X (1.23 Parameters)

DCM (for Dose X X X X projections)

Class A Model (to X X Plume Exposure EPZ)

~

Class B Model or X X Class A Model (to Ingestion EPZ)

(6) a. July 1, 1982 or at the time of the completion of milestone 5, whichever is sooner.

b. Mandatory review of the DCM by the licensee.

c. Any DCM in use should be reviewed to ensure consistency with the operational Class A model. Thus, actions recommended during the initial phases of a radiological emergency would

'e consistent with those after the TSC and EOF are activated.

(7) a. September 1, 1982.

b. Description of the Class B model provided to the NRC.

c. Documentation of the technical bases and justification for selection of the type Class B model by the licensee with a discussion of the site-specific attributes.

(8) a. June 1, 1983.

b. Full operational capability of the Class B model.

c. Class B model of element 3 of Appendix 2 to NUREG-0654, Revision 1 Applicants for an operating license shall meet at least milestones 1, 2, and 3 prior to the issuance of an operating license. Subsequent milestones shall be met by the same dates indicated for operating reactors. For the alternative to milestone 3, the meteorological measurements program shall be consistent with the NUREG-75/087, "Standard Review Plan for the Review of Safety Analysis Reports for X.1-62

Nuclear Power Plants," Secton 2.3.3 program as the baseline or element 1 and/or element 2 systems.

X.l.68.2 Inter retation None required.

X.l.68.3 Statement of Res onse Milestones 1, 2 and 3 are beirrg addressed as a part of the short term emergency preparedness requirement III.A.l.l: Refer to subsection X.2.38 for'esponse. No other milestones require a response at this time.

X.l.69 INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT LIKELY TO CONTAIN RADIOACTIVE MATERIAL (III.D.l.l)

X.1.69.1 Statement of Re uirement Applicants shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels. This program shall include the following:

(1) Immediate leak reduction.

(a) Implement all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment.

(b) Measure actual leakage rates with system in operation and report them to the NRC.

(2) Continuing Leak Reduction--Establish and implement a program of preventive maintenance to reduce leakage to as-low-as-practical levels. This program shall include. periodic integrated leak tests at intervals not to exceed each refueling cycle'.

This requirement shall be implemented prior to issuance of a'ull-power license.

Applicants shall provide,a summary description, together with initial leak- .

test results, of their program to reduce leakage from systems outside containment that would or could contain primary coolant or other highly radioactive fluids or gases during or following a serious transient or accident. Applicants shall submit this information at least 4 months prior to fuel load.

X.l.69.2 Inte retation None required.

X.l.69.3 Statement of Res onse A program to reduce'leakage will be developed by fuel load. This program will include the following:

(1) Identification of all applicable flow paths through containment penetrations excluding instrumentation tubes, electrical conduits, and manways.

(2) Evaluation of current leak detection systems and comparison to operating plants with designs similiar to SSES.

(3) Develop procedures.

Appropriate documentation will be made available for review by Region I I&E.'he program will implemented prior to issuance of a full power license.

X.1.70 INPLANT IODINE RADIATION HONITORING (III.D.3.3)

X.l.70.1 Statement of Re uirement Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident.

Each applicant for a fuel-loading license to be issued prior to January 1, 1981 shall provide the equipment, training, and procedures necessary to accurately determine the presence of airborne radioiodine in areas within the plant where plant personnel may be present during an accident.

Effective monitoring of increasing iodine levels in the buildings under accident conditions must include the use of portable instruments using sample media that will collect iodine selectively over xenon (e.g., silver zeolite) for the following reasons:

(1) The physical size of the auxiliary and/or fuel handling building precludes locating stationary monitoring instrumentation at all areas where airborne iodine concentration data might be required.

(2) Unanticipated isolated "hot spots" may occur in locations where no stationary monitoring instrumentation is located, X.l-64

(3) Unexpectedly high background radiation levels near stationary monitoring instrumentation after an accident may interfere with filter radiation readings.

(4) The time required to retrieve samples after an accident may result in high personnel exposures if these filters are located in high-dose-,rate areas.

After January 1, 1981, each applicant and licensee shall have the capability to remove the sampling cartridge to a low-background, low-contamination area for further analysis. Nor'mally, counting rooms in auxiliary buildings will not have sufficiently low backgrounds for such analyses following an accident. In the low background area, the sample should first be purged of any entrapped noble gases using nitrogen gas or clean air free of noble gases. The licensee shall have the capability to measure accurately the iodine concentrations present on these samples under accident conditions. There should be sufficient samplers to sample all vital areas.

For applicants with fuel-loading dates prior to January 1, 1981, provide at fuel loading (until January 1, 1981) the capability to accurately detect the presence of iodine in the region of interest following an accident.

This can be accomplished by using a portable or cart-mounted iodine sampler

~

with attached single-channel analyzer (SCA). The SCA window should be calibrated to the 365 keV of iodine-131 using the SCA. This will give an initial conservative estimate of presence of iodine-131 using the SCA.

This will give an initial conservative estimate of presence of iodine and can be used to determine if respiratory protection is required. Care must be taken to assure that the counting system is not saturated as a result of too much activity collected on the sampling cartridge. Applicants shall meet these requirements prior to fuel load.

X.l.70.2 Inte retation PPGL is in basic agreement with the technical discussion as outlined in this requirement. It should be noted that SSES is a BVR and does not possess an auxiliary building. Consequently, it is premature to suggest that our count'ing facilities within the control structure will be inadequate to effectively count air samples. Additionally, purging of the air sample cartridges may not be necessary if an effective collection media is used for radioiodine air sampling.

X.l.70.3 Statement of Res onse Two approaches are being developed to monitor radioiodine levels in areas where plant personnel may be during an accident. The first involves the use of portable equipment capable of collecting radioiodine selectively, and measuring samples with a two channel analyzer. Alternately, the sample cartridge will be taken.to a low background counting area. Then sample preparation procedures that reduce the noble gas contribution will be

utilized and activity measured by a multichannel analyzer/detection system.

Procedures and equipment will be available for use by fuel load.

Appropriate documentation will be made available for review by Region I I&E.

X.l.71 CONTROL ROOM HABITABILITYRE(}UIREMENTS (III.D.3.4)

X.1.71.1 Statement of Re uirement Licensees shall assure that control room operators will be adequately protected against the effects of accidental release of toxic and radioactive gases and that the nuclear power plant can be safetly operated or shut down under design basis accident conditions (Criterion 19, "Control Room," of Appendix A, "General Design Criteria for Nuclear Power Plants,"

to 10 CFR Part 50).

All licensees must make a submittal to the NRC regardless of whethe'r or not they met the criteria of the Standard Review Plans (SRP) sections listed below. The new clarification specifies that licensees that meet the criteria of the SRPs should provide the basis by referencing past submittals to the.NRC and/or providing new or additional information to supplement past submittals.

X.1.71.1.1 Re uirements for Licensees that Meet Criteria All licensees with control rooms that meet the criteria of the following sections of the Standard Review Plan:

2.2.1-2.2.2 Identification of Potential Hazards in Site Vicinity 2.2.3 Evaluation of Potential Accidents; 6:4 Habitability Systems shall report their findings regarding the specific SRP sections as explained below. The following documents should be used for guidance:

(a) Regulatory Guide 1.78, "Assumptions for Evaluating the Habitability of Regulatory Power Plant Control Room During a Postulated Hazardous Chemical Release";

(b) Regulatory Guide 1.95, "Protection of Nuclear Power Plant Control Room Operators Against an Accident Chlorine Release"; and, (c) K. G. Murphy.and K. M. Campe, "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criterion 19,"

13th AEC Air Cleaning Conference, August 1974.

Licensees shall submit the results of their findings as well as the basis for those findings by January 1, 1981. In providing the basis for the habitability finding, licensees may reference their past submittals.

Licensees should, however, ensure that these submittals reflect the. current

'.1-66

facility design and that the information requested in Attachment 1 of NUREG 0737 is provided.

E X.l.71.1;2 Re uirements for Licensees that Do Not Meet Criteria All licensees with control rooms that do, not meet the criteria of the above-listed references, Standard Review Plans,'egulatory Guides, and other references shall perform the evaluations and identify appropriate modifications, as discussed below.

Each licensee submittal shall include the results of the analyses of control room concentrations from postulated accidental release of toxic gases and control room operator radiation "exposures from airborne radioactive material and direct radiation resulting from design-basis accidents. The toxic gas accident analysis should be performed for all potential hazardous chemical releases occurring either on the site or within 5 miles of the plant-site boundary. Regulatory Guide 1.78 lists the chemicals most commonly encountered in the evaluation of control room habitability but is not all inclusive.

The design-basis-accident (DBA) radiation source term should be for the loss-of-coolant accident LOCA containment leakage and engineered safety feature (ESF) leakage contribution outside'ontainment as described in Appendix A and B of Standard Review Plan Chapter 15.6.5. In addition, boiling-water reactor (BWR) facility evaluations should add any leakage from the main steam isoaltion valves (MSIV) "(i.e., valve-stem leakage, valve seat leakage, main steam isolation valve leakage control system release) to the containment leakage and ESF leakage following a LOCA. This should not be construed as altering the staff recommendations in Section D of Regulatory Guide 1.96 (Rev. 2) regarding MSIV leakage-control systems.

Other DBAs should be reviewed to determine whether they might constitute a more-severe control-room hazard than the LOCA.

In addition to the accident-analysis results, which should either identify the possible need for control-room modifications or provide assurance that the habitability systems will operate under all postulated conditions to permit the control-room operators to remain in the control room'o take appropriate actions required by General Design Criterion 19, the licensee should submit sufficient information needed for an independent evaluation of the adequacy of the habitability systems'. Attachment 1 of NUREG 0737 lists the information that should be provided along with the licensee's evaluation.

Applicants for operating licenses shall submit their responses prior to issuance of a full-power license. Modifications needed for compliance with the control-room habitability requirements specified in this letter should be identified, and a schedule for completion of the modifications should be provided. Implementation of such modifications should be started without awaiting the results of the staff review. Additional needed modifications, X.1-67

if any, identified by the staff during its review will be specified to licensees.

X.l.71.2 Inter retation None required.

X.l.71.3 Statement of Res onse.

The control room HVAC system Nyout and functional design includes of the control room from radioactive and toxic gases. 'rotection Subsection 6.4 provides a complete description of this system.and compliance to habitability requirements. Refer to subsection 6.4 for the response to this requirement. (This subsection is being revised to include the response).

X.1-68

X.2 RESPONSE TO RE UIREMENTS IN NUREG 0694 NUREG 0694 supersedes NUREG 0578. The clarifications given in the Vassallo letter on November 9, 1979 were used in the development of applicable responses.

X.2.1 SHIFT TECHNICAL ADVISOR (I.A.l.l)

Requirement superseded by NUREG 0737. Refer to Subsection X.l.l for response.

K.2.2 SHIFT SUPERVISOR ADMINISTRATIVE DUTIES (I.A.1.2)

X.2.2.1 .Statement of Re uirement Review the administrative duties of the shift supervisor and delegate functions that detract from or are subordinate to the management responsibility for assuring safe operation of the plant to other personnel not on duty in the control room. This requirement shall be met before fuel load.

X.2.2.2 Inter retation None required.

X.2.2.3 Statement of Res onse PPGL has restructured the operations organization and redefined responsibilities of shift personnel to relieve the shift supervisor of routine administrative duties. The job descriptions of all shift personnel have been revised. Appropriate documentation will be made available for review by Region I I&E.

X.2.3 SHIFT MANNING (I.A.1.3)

Requirement superseded by NUREG 0737. Refer to Subsection X.1.3 for response.

X.2.4 IMMEDIATE UPGRADING OF OPERATOR AND SENIOR OPERATOR TRAINING AND UALIFICATION (I.A.2.1)

Requirement superseded by NUREG 0737. Refer to Subsection X.1.4 for response.

X.2.5 REVISE SCOPE AND CRITERIA FOR IICENSING EXAMINATIONS (I.A.3.1)

Requirement superseded by NUREG 0737. Refer to Subsection X.l.6 for response.

X.2.6 EVALUATION OF ORGANIZATION AND MANAGEMENT IMPROVEMENTS OF NEAR-TERM OPERATING LICENSE APPLICANTS (I.B.1.2)

X.2.6.1 Statement of Re uirement The licensee organization shall comply with the findings and requirements generated in an interoffice NRC review of licensee organization and management. The review will be based on an NRC document. entitled Draft Criteria for Utility Management and Technical Competence. The first draft of this document was dated February 25, 1980, but the document is changing with use and experience in ongoing reviews. These draft criteria address the organization, resources, training, and qualifications of plant staff, and management (both onsite and offsite) for routine operations and the resources and activities (both onsite and offsite) for accident requirement shall be met prior to fuel load.

conditions'his X.2.6.2 Inter retation None required.

X.2.6.3 Statement of Res onse A review of organization and management has been completed in accordance with draft NUREG 0731, "Guidelines for Utility Management Structure and Technical Competence." The results of this review will serve as a basis for the NRC audit of the organization which is scheduled for March 1981.

A schedule for responding to recommendations made during the audit will be developed prior to fuel load.

X.2.7 SHORT TERM ACCIDENT ANALYSIS AND PROCEDURE REVISION (I.C.1)

Requirement superseded by NUREG 0737. Refer to Subsection X.l.8 for response.

X.2.8 SHIFT RELIEF AND TURNOVER PROCEDURES (I.C.2)

X.2.8.1 Statement of Re uirement Revise plant procedures for shift relief and turnover to require signed checklists and logs to assure that the operating staff (including auxiliary operators and maintenance personnel) possess adequate knowledge of critical X. 2-2

plant parameter status, system status, availability and alignment. This requirement shall be met prior to fuel load.

X.2.8.2 Inter retation None required.

X.2.8.3 Statement of Res onse The procedures necessary to meet this requirement are being prepared and will be implemented prior to fuel load. These procedures will require the use of checklists as stated above. Appropriate documentation will be made available for review by Region I ISZ.

X.2.9 .SHIFT SUPERVISOR RESPONSIBILITIES (I.C.3)

X.2.9.1 Statement of Re uirement Issue a corporate management directive that clearly establishes the command duties of the shift supervisor and emphasizes the primary management responsibility for safe operation of the plant. Revise plant procedures to clearly define the duties, responsibilitie's and authority of the shift supervisor and the control room operators. This requirement-shall be met "

prior to fuel load.

X.2.9.2 Inter retation None required.

X.2.9.3 Statement of Res onse The job descriptions of all shift personnel have been revised and administrative procedures will be reviewed and revised in compliance with this requirement. A corporate directive will be issued to establish the command duties of the shift supervisor. All tasks will be completed prior to fuel load. Appropriate documentation will be made available for review by Region I ISE.

X.2.10 CONTROL ROON ACCESS (I.C.4)

X.2.10.1 Statement of Re uirement Revise plant procedures to limit access to the control room to those individuals responsible for the direct operation of the plant, technical advisors, specified NRC personnel, and to establfsh a clear line of authority, responsibility, and succession in the control room. This requirement shall be met prior to fuel load.

X.2-3

X.2.10.2 Inte retation None required.

X.2.10.3 Statement of Res onse Administrative procedures and job descriptions of all shift personnel will be revised priot to fuel load to establish the chain of command and address control room access. Appropriate documentation will be made available for review by Region I ISE.

X.2.11 PROCEDURES FOR FEEDBACK OF OPERATING EXPERIENCE TO PLANT STAFF I.C.5 Requirement superseded by NUREG 0737. Refer to Subsection X.l.12 for response.

X.2.12 NSSS VENDOR REVIEW OF PROCEDURES (I.C.7)

X.2.12.1 Statement of Re uirement Obtain nuclear steam supply system vendor review of low-power testing procedures to further verify their adequacy. This requirement shall be met prior to fuel load.

Obtain NSSS vendor review of power-ascension test and emergency procedures to further verify their adequacy. This requirement must be met before issuance of a full-power license.

X.2.12.2 Inter retation None required.

X.2.12.3 Statement of Res onse The General Electric Company will review the low-power testing, power ascension, and emergency procedures. The review of emergency procedures will consider the Emergency Procedure Guidelines submitted to the NRC by letter from R. H. Buchholz to D. G. Eisenhut on June 30, 1980. All reviews will be completed by fuel load.

X.2-4

K.2.13 PILOT MONITORING OF SELECTED EMERGENCY PROCEDURES FOR NEAR-TERM OPERATING LICENSE APPIICANTS (I.C.8)

'I X.2.13.1 Statement of Re uirement Correct emergency procedures, as nenes'sary, based on the NRC audit of selected plant emergency operating procedures (e.g., small-break LOCA, loss of feedwater, restart of engineered safety features following a loss of AC power or, steam-line break).

X.2.13.2 Inter retation None required.

X.2.13.3 Statement of Res onse No response is necessary until the NRC completes the audit and issues specific requirements.

X.2.14 CONTROL ROOM DESIGN (I.D.1)

Requirement superseded by NUREG 0737. Refer to Subsection X.1.16 for response.

X.2.15 TRAINING DURING IOW POWER TESTING (I.G.I)

X.2.15.1 Statement of Re uirement Define and commit to a special low-power testing program approved by NRC to be conducted at power levels no greater than 5 percent for the purposes of providing meaningful technical information beyond that obtained in the normal startup test program and to provide supplemental training. This.

requirement shall be met before fuel load.

Supplement operator training by completing the special low-power test

'rogram. Tests may be observed by other shifts or repeated on other shifts to provide training to the operators. This requirement shall be met before issuance of a full-power license.

X.2.15.2 Inter retation, None required.

X.2.15.3 Statement of Res onse A special low power testing program will be conducted following NRC approval of the program that will be submitted by the BWR Owners'roup.

Augmented operator training will be maximized by appropriate scheduling of tests and performance of tests on the simulator.

X.2.16 . REACTOR COOLANT SYSTEM VENTS (II.B.l)

Requirement superseded by NUREG 0737. Refer to Subsection X.1.19 for response X.2.17 PLANT SHIELDING (II.B.2)

Requirement superseded response.

by NUREG 0737 'efer to Subsection X.l.20 for X.2.18 POSTACCIDENT SAMPLING (II.B.3)

Requirement superseded by NUREG 0737. Refer to Subsection X.l.21 for response.

'.2.19 TRAINING FOR MITIGATING CORE DAMAGE (II.B.4)

Requirement superseded by NUREG 0737. Refer to Subsection X.1.22 for response.

X.2.20 RELIEF AND SAFETY VALVE TEST RE(}UIREMENTS (II.D.1)

Requirement superseded by NUREG 0737. Refer to Subsection X.1.23 for response.

X.2.21 RELIEF AND SAFETY VALVE POSITION INDICATION (II.D.3)

Requirement superseded by NUREG 0737. Refer to Subsection X.1.24 for response.

X.2.22 CONTAIRiENT ISOLATION DEPENDABILITY (II.E.4.2)

Requirement superseded by NUREG 0737. Refer to Subsection X.l.29 for response.

K.2.23 ADDITIONAL ACCIDENT MONITORING INSTREKNTATION (II.F.1)

Requirement superseded by NUREG 0737. Refer to Subsection X.l.30 for response.

X.2-6

X.2.24 INADEQUATE CORE COOLING INSTREKNTS (II.F.2)

Requirement superseded by NUREG 0737. Refer to Subsection X.l.31 for response; X.2.25 ASSURANCE OF PROPER ESF FUNCTIONING (II.K.1.5)

X.2.25.1 Statement of Re uirement Review all valve positions, positioning requirements, positive controls and related test and maintenance procedures to assure proper ESF functioning.

This requirement shall be met by fuel load.

X.2.25.2 Inter retation

'I None required. 'I X.2.25.3 Statement of Res onse As test and maintenance procedures are developed, they will be reviewed to insure safety related valve position concerns aie considered. Appropriate documentation will be made available for review by Region I ISE.

X.2.26 SAFETY RELATED SYSTEH OPERABILITY 'STATUS (II.K.1.10)

X.2.26.1 Statement of Re uirement Review and modify", as required, procedures for removing safety-related systems from service (and restoring to service) to assure operability status is known. This requirement. shall be met by fuel load.

X.2.26.2 Inter retation None required.

X '.26.3 Statement of Res onse As test and maintenance procedures are developed they will be reviewed to insure safety system operability status concerns are considered.

Appropriate documentation will be made available for review by Region I ISE.

X.2.27 TRIP PRESSURIZER LOW-LEVEL COINCIDENT SIGNAL BISTABLES (II.K.1.17)

This requirement is not applicable to SSES.

X.2.28 OPERATOR TRAINING FOR PROMPT MANUAL REACTOR TRIP (II.K.1.20)

This requirement is not applicable to SSES.

X.2,29 AUTOMATIC SAFETY GRADE ANTICIPATORY TRIP (II.K.1.21)

This requir'ement is not applicable to SSES.

X.2.30 AUXILIARYHEAT REMOVAL SYSTEMS OPERATING PROCEDURES (II.K.1.22)

X.2.30.1 Statement of Re uirement Describe the automatic and manual actions necessary for proper functioning of the auxiliary heat removal systems that are used when the main feedwater system is not operable. This requirement shall be met by fuel load; X.2.30.2 Inter retation None required.

X.2.30.3 Statement of Res onse The response to this requirement was provided by General Electric in NEDO-24708, "Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors," (August 1979) and supplement I.'.2.31 REACTOR LEVEL INSTRUMENTATION (II.K.1.23)

X.2.31.1 Statement of Re uirement For boiling water reactors, describe all uses and types of reactor vessel level indication for both automatic and manual initiation of safety systems. Describe other instrumentation that might give the operator the same information on plant status. This requirement shall be met before fuel load:

X.2.31.2 Inter retation None required.

X.2.31.3 Statement of Res onse The response to this requirement was provided by General Electric in.NEDO-24708, Additional Information Required for NRC Staff G'eneric Report on Boiling Water Reactors," (August 1979) and Supplement I.

X.2-8

X.2.32 COMMISSION ORDERS ON BABCOCK AND WILCOX PLANTS (II.K.2)

These requirements are not applicable to SSES.

X.2.33 REPORTING REQUIREMENTS FOR SAFETY/RELIEF VALVE FAILURES OR CHALLENGES II.K.3.3 X.2.33.1 Statement .of Re uirement Assure that any failure of a'PORV or safety valve to close will be reported to the NRC promptly. All challenges to the PORVs or safety valves should be documented in the annual report. This:requirement shall be met before issuance of a full-power license.

X.2.33.2 .Inter retation None required.

X.2.33.3 Statement of Res onse Section 6.9.1.6 of the technical specifications requires safety/relief valve (SRV) challenges to be reported in the monthly operating report. The technical specifications will be updated by fuel load to identify reporting requirements for SRV failures.

X.2.34 PROPORTIONAL INTEGRAL DERIVATIVE CONTROLLER (II.K.3.9)

This requirement is not applicable to SSES.

X.2.35 ANTICIPATORY REACTOR TRIP MODIFICATION (II.K.3.10)

This requirement is not applicable to SSES.

X.2.36 POWER OPERATED RELIEF VALVE FAILURE RATE (II.K.3.11)

This requirement is not applicable to SSES ~

X.2.37 ANTICIPATORY REACTOR TRIP ON TURBINE TRIP (II.K.3.12)

This requirement is not applicable to SSES.

X.2-9

X.2.38 EMERGENCY PREPAREDNESS-SHORT TERM (III.A.l.l)

X.2.38.1 Statement of Re uirement Comply with Appendix E, "Emergency Facilities," to 10 CFR Part 50, Regulatory Guide 1.101, ."Emergency Planning for Nuclear Power Plants," and for the offsite plan's, meet essential elements of NUREG-75/ill (Ref. 28) or have a favorable finding from FEMA. This requirement shall be met prior to fuel load.

Provide an emergency response plan in substantial compliance with NUREG'-

0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants" (which may be modified as a result of public comments solicited in early 1980) except that only a description of and completion schedule for the means for providing p'rompt notification to the population (App. 3), the staffing for emergencies in addition to that already required (Table B.l), and an upgraded meteorological program (App. 2) need be provided (Ref. 10). NRC will give substantial weight findings on offsite plans in judging the adequacy against NUREG-0654, Perform an emergency response exercise to test the integiated capability and a major portion of the basic elements existing within emergency preparedness plans and organizations. This requirement shall be met before issuance of a full-power license.

X.2.38.2 Inter retation PPGL is interpieting Emergency Facilities as encompassing'hose requirements for TSC, Interim TSC, EOF, Interim EOF, SPDS, OSC as outlined in draft NUREG 0696 and TMI Action Items in 0737. Complete Site, State, County, Township and Municipality Emergency Plans using the Guidelines of NUREG-0654 Rev. 1. Exercise the plans to ensure they are integrated and workable, Comply with meteorological requirements of NUREG 0654 Appendix 2, Rev. 1 X.2.38.3 Statement of Res onse Emergency facility design criteria was submitted to NRC for review and approval in January 1981. Interim facility use will be so defined in the SSES Emergency Plan and comply with NUREG 0737 requirements. This- complies with NRC schedules and requirements as outlined in draft NUREG 0696.

SSES Emeigency Plan Rev. 2 was submitted to the -NRC 10/30/80 complying with 10 CFR 50 Appendix, E and General Criteria of draft NUREG 0654. NRC comments have not been received. SSES Emergency Plan Rev. 3 complying with NRC comment letter and NUREG 0654 Rev. 1 will be submitted subsequent to receipt of NRC comments. Commonwealth of Pennsylvania Emergency Plan, has been submitted to FEMA complying with NUREG 0654 Rev. i. County, Township .

and Municipality Plans are scheduled for submittal to FEMA by 3/1/81.

SSES Integrated Emergency Exercise will be held prior to fuel load.

l

The SSES Emergency Plan, Rev. 3 and the SSES Emergency Plan Implementing Procedures will meet the requirements of NUREG 0654, Appendix 2 with one exception. The computerized dose projection calculations will be based on a straight-line Gaussian model with weather and building wake correction factors included in the methodology. The existing computer system does not possess the core space or computing power to perform sophisticated calculations employing, site specific terrain correction factors. It is PPGL"s intent to have a computer model that employs site specific terrain correction factors when the Emergency Response Computer System (associated, with the Emergency Operations Facility). is installed.

K.2.39 UPGRADE EMERGENCY SUPPORT FACILITIES (III.A.1.2)

Requirement superseded by NUREG 0737. Refer to Subsection X.1.67 for response.

X.2.40 PRIMARY COOLANT SOURCES OUTSIDE CONTAINMENT (III.D.1.1)

Requirement superseded by NUREG 0737. Refer to Subsection X.l.69 for response.

X.2.41 INPLANT RADIATION MONITORING (III..D.3.3)

Requirement superseded by NUREG 073?. Refer to Subsection X.l.70 for response.

X.2.42

~ ~ CONTROL ROOM HABITABILITY (III.D.3.4)

Requirement superseded by NUREG 0737. Refer to Subsection X.1.71 for response.

X.2-11

ML18026A221

Text

Dear Mr. Youngblood:

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