ML19340D164

Forwards Response to Applicable Requirements of NUREG-0737 Re TMI Action Plan.Progress of Response Impeded Due to Lack of Guidance Re Applicability of Requirements to Gas Cooled Reactor

ML19340D164
Person / Time
Site:	Fort Saint Vrain
Issue date:	12/20/1980
From:	Warembourg D PUBLIC SERVICE CO. OF COLORADO
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.3, TASK-1.A.2.3, TASK-1.A.3.1, TASK-2.F.1, TASK-TM P-80438, TAC-46448, TAC-57944, NUDOCS 8012290231
Download: ML19340D164 (69)

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December 20, i;80 Fort St. Vrain Unit No. 1 P-80438 J

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Mr. Darrel G. Eisenhut, Director Division of Reactor Licensing Office of Nuclear Reactor Regulation g

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SUBJECT:

Fort St. Vrain Unit No.1 TMP Action Plan Requirements NUREG 0737

REFERENCE:

NRC Letter Dated 10-31-30

Dear Mr. Eisenhut:

We have completed our review of the subject NUREG transmitted by the above referenced letter. The attachment contains our response to each of the action requirements that are applicable to Fort St.

Vrain. Our response to the various action requirements generally falls into four (4) categories.

1.

Those action requirements for which we have provided previous response which we feel is still applicable in light of the clarification provided by NUREG 0737. Other than previous ccmmitments that may have been made as a part of our response, we do not plan on any further action.

2.

Those action requirements and schedules which for various reasons we will be unable to meet, or which for various reasons we have taken exception as to the applicability of the requirements to gas cooled technology as opposed to water cooled technology.

3.

Those action requirements and schedules which we intend to meet.

4.

Those action requirements which are clearly not applicable to Fort St. Vrain.

As we have pointed out in previous correspondence we have had a Nj difficult time applying the criteria, guidance and requirements to Fort St.

Vrain, and in many cases have had little if any guidance 3

that was clearly applicable to gas cooled technology.

In addition, we were consistently exciuded from receipt of various letters, bulletins, and orders resulting from the TMI action requirements, and

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in this respect, we find that we were not afforded the same time schedule to plan and complete various activities by comparison to the water reactors, so m go m p

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We met with the Special Projects Division along with other members of the staff on December 10, 1980, in an attempt to obtain some further clarification of the applicability of many of the action requirements to gas cooled technology. While we were able to obtain some general clarification we found that we were unable to obtain any clarification with reference to the technical differences involving gas cooled reactors versus water cooled reactors which is a continuing problem.that has been with us since the onset of TMI, and which because of a lack of guidance, has impeded our progress in many areas.

We have continued in our efforts to justify certain exceptions from the various criteria on the basis of distinct differences between a gas cooled reactor and a water cooled reactor. We believe that adequate technical justification has been provided in many areas, but it is obvious that the technical justification provided is not being considered in the various staff reviews, and it is also obvious that various technical justifications which were reviewed by one group in the past are not being considered as new review groups are formed.

As a result we appear to be in a continual education process and in our opinion, continue to be penalized with the inapplicable criteria.

Very truly yours,

$ W WA Don W. Warembourg

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Manager, Nuclear Production Fort St. Vrain Nuclear Generating Station DWW/alk Attachment l

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ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 I.A.1.1 Shift Technical Advisor Action Requirement:

No later than January 1,-1981, all licensees of operating reactors shall provide this office with a description of their STA training program and their plans for requalification training.

This description shall indicate the level of training attained by STA's by January 1, 1981, and demonstrate conformance with the qualification and training requirements in the October 30, 1979, letter.

Applicants for operating licenses shall provide the same information in their application, or a.mendments thereto, on a schedule consistent with the NRC licensing review schedule.

No later than January 1, 1981, all licensees of operating reactors shall provide this office with a description of their long-term STA program, including qualification, selection criteria, training plans, and plans, if any, for the eventual phaseout of the STA program.

(NOTE: The description shall include a

comparison of the licensee / applicant program with the above-mentioned INPO document. This request solicits industry views to assist NRC in establishing long-term improvements in the STA program. Applicants for operating licenses shall provide the same information in their application, or amendments thereto, on a schedule consistent with the NRC licensing review schedule.)

(1) Training that meets the lessons-learned requirements shall be completed by January 1, 1981, or by the time the fuel-loading license is issued, whichever is later.

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. ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action plan Requirements NUREG 0737 (2) A description of the current training program and demonstration of conformance with the October 30, 1979, letter shall be submitted:

(a) No later than January 1, 1981, for licensees of operating reactors; and, (b) On a schedule consistent with review schedule for applicants for operating licenses.

(3) A description of the long-term STA program shall be submitted:

(a) No later than January 1, 1981, for licensees of operating reactors; and, (b) On a schedule consistent with review schedule for applicants for operating licenses.

PSC Resoonse:

(1) With reference to Item (1) training of the Technical Advisors has been completed.

(2) A description of the training program which has been completed will be submitted by January 1, 1981.

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(3) A description of the long term Technical Advisor training program will be submitted by January 1, 1981.

I.A.1.3 Shift Manning Action Reauirement:

(1) Overtime administrative procedures shall be established for operating reactors by November 1, 1980, and by fuel loading for applicants for operating licenses.

Staffing requirements shall be completed by July 1, 1982, for operating reactors and by fuel load for operating license applicants.

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i ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 i

PSC Response:

(1) Based on your letter of July 31, 1980, overtime administrative procedures were established by November 1, 1980, as outlined in our letter P-80388.

Although NUREG 0737 made some modification to the work schedule we feel our response is still applicable. As we indicated in P-80388 we did take a minor exception to the guidance involving work for more than 14 consecutive days and requested 16 consecutive days to accomodate our shift rotation cycle.

In the 4

December 10, 1980, meeting in Bethesda with members of the staff it was indicated that this exception, based on a preliminary review, would be acceptable.

(2) We plan to meet the minimum staffing requirements by July, 1982.

I.A.2.1 Immediate Upgrade, Reactor Operator and Senior Reactor Operator Training and Qualifications Action Requirement:

Effective December 1, 1980, an applicant for a senior reactor operator (SRO) license will be required to have been a licensed operator for 1 year.

i This requirement applies to applicants for senior reactor operator licenses received after December 1, 1980.

(1) Our SRO program has been upgraded to meet the requirements setforth.

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- ATTACHMENT 1 Fort St. Vrain Unit No. 1 Resoonse Clarification of TMI Action Plan Requirements NUREG 0737 (2) Concerning qualifications of reactor operators we never received the March 28, 1980, letter until it was issued as a part of the preliminary draft to NUREG 0737. We did not become aware of the existance of this letter in sufficient time to permit us to submit a program for revised R0 training and instructor training by August 1, 1980, as required by Enclosure 1 of the March 28, 1980, letter. We have been able to upgrade our program and we were able to get our training instructor license applications in prior to August 1, 1980.

We will submit our programs to OLB by January 15, 1981.

I.A.2.3 Administration of Training Programs Action Requirement:

Pending accreditation of training institutions, licensees and applicants for operating licenses will assure that t.aining center and facility instructors who teach systems, integrated responses, transient, and simulator courses demonstrate senior reactor operator (SRO) qualifications and be enrolled in appropriate requalification programs.

The requirements for operating reactors have been completed.

Applications for SRO examinations should be submitted.

All applicants for operating license should submit documentation 2 months prior to the expected issuance of an operating license.

pSC Resoonse:

We have revised our program to meet the requirements setforth. As indicated in our response to Item I.A.2.1 our program will be submitted by January 15, 1981.

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...... ATTACHMENT 1 Fort St. Vrain Unit No. I Response Clarification of TMI Action Plan Requirements NUREG 0737 I.A.3.1 Revise Scope and Criteria for Licensing Examinations Simulator Exam (Item 3)

Action Requirement:

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

The schedule for operating reactors is October 1,1981, for licensees without simulaters and June,1980, for licensees with simulators.

The schedule for applicatns for operating license without simulators is October 1, 1981, or prior to fuel

load, whichever is later, including cold examinations.

The schedule for applicants for operating license with simulators is prior to full load including cold examination.

PSC Response:

On December 10, 1980, we met with the staff to discuss simulator training and the applicability of the October 31, 1981, date for simulator training. The basis of the matter is that we are a unique reactor and there are no simulator training facilities either site specific or generic that we can utilize to meet the simulator training requirement.

It would be of no value to subject any of our operators to existing simulator training that has been developed for water reactors. On this basis we proposed to the staff that we could upgrade our operator training program to concentrate more on hands-on operating experience and on more training in the area of transients.

The staff indicated that the requirement for simulator training was presently being proposed for inclusion in 10CFR50.55 and they were not in a position to grant any relief in this area.

It is obvious that we cannot meet an October date for simulator training and it is not reasonable to apply a simulator requirement on a one-of-a-kind plant.

On this basis it is our intent to appeal this requirement to the Commission.

_ ATTACHMENT 1 Fort St. Vrain Unit No.1 Response Clarification of TMI Action Plan Requirements NUREG 0737 I.B.I.2 Indeoendent Safety Engineering Group Requirements are not applicable to Fort St. Vrain.

I.C.1 Guidance for the Evaluation and Development of Procedures for Transients and Accidents Action Recuirement:

In letters of September 13 and 27, October 10 and 30, and November 9, 1979, the Office of Nuclear Reactor Regulation required licensees of operating plants, applicants for operating licenses and licensees of plants under construction to perform analyses of transients and accidents, prepare emergency procedure guidelines, upgrade emergency procedures, including procedures for operating with natural circulation conditions, and to conduct operator retraining (see also item I.A.2.1).

Emergency procedures are required to be consistent with the actions necessary to cope with the transients and accidents analyzed. Analyses of transients and accidents were to be completed in early 1980 and implementation of procedures and retraining were to be completed 3 months after emergency procedure guidelines were established; however, some difficulty in completing these requirements has been experienced.

Clarification of the scope of the task and appropriate schedule revisions are being developed.

In the course of review of these matters on Babcock and Wilcox (B&W)-designed plants, the staff will follow up on the bulletin and orders matters relating to analysis methods and results, as listed in NUREG 0660, Appendix C (see Table C.1, items 3, 4, 16,-18, 24, 25, 26, 27; Table C.2, items 4, 12, 17, 18, 19, 20; and Table C.3, items 6, 35, 37, 38, 39, 41, 47, 55, 57).

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_ _ _. ATTACHMENT 1 Fort St. Vrain Unit No.1 Response Clarification of TMI Action Plan Requirements NUREG 0737 PSC Response:

While Clarification Item I.C.1 appears to be specifically directed to LWR's, PSC considers its intent also applicable to HTGR's (Fort St.

Vrain).

PSC confi rms that the applicable requirements of Clarification Item I.C.1 have been implemented at the Fort St. Vrain Nuclear Generating Station.

The original FSAR accident analyses and later FSAR accident reanalyses have been used to develop and revise the Fort St.

Vrain Emergency Procedures.

These analyses have been periodically submitted to NRC for review, e.g. the two most recent reanalyses requiring procedure changes being:

(1) Permanent Loss of Forced Circulation (LOFC) - The FSAR Design Basis Accident No. I reanalyses, which requires rapid depressurization of the PCRV within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following an LOFC and simultaneous Loss of PCRV Cooling Water for a period of time, was submitted to NRC by PSC letter P-77221.

In order to maintain the FSAR condition of 1,500 degrees Fahrenheit liner temperature at 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> into the LOFC and to assure LTA performance, the emergency procedures have been revised to initiate rapid depressurization of the PCRV as early as 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> l

following an LOFC depending upon the average core outlet temperature at the onset of a LOFC.

This analysis, which was (2) Manual Valve Study submitted by PSC letter P-78071, considered the emergency procedure changes necessary to ensure that the firewater cooling system is available for the helium circulator Pelton wheels and the steam l

generators.

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! ATTACHMENT 1 Fort St. Vrain Unit No. I Response Clarification of TMI Action Plan Requirements NUREG 0737 PSC considers that the current Fort St. Vrain " Emergency Procedures" coupled with the " Abnormal Procedures for Shutdown Cooling" and " Safe Shutdown and Cooling with Highly Degraded Plant Conditions Procedures" meet the intent of I.C.1 requirements, since these procedures were developed from the original FSAR accident analyses and later FSAR accident reanalyses. These Emergency Procedures, which were developed from plant-specific accident analyses into SYMPTON-IMMEDIATE ACTION-F0LLOWUP ACTION formatted procedures (somewhat like the BWR Owner's Group Guidelines referred to by Clarification Item I.C.1), simply direct (or guide) the reactor operators to regain water flow through the steam generator (s) and helium flow through the reactor i

core in order to mitigate the consequences of any failure or accident that could result in inadequate core cooling.

Therefore, the current Emergency Procedures are considered to be not only Guidelines but also the procedures developed from the " Guidelines" previously ~ submitted to NRC.

For example, the previously mentioned Valve Study Analysis which was submitted by P-78071 included revisions to the Fort St.

Vrain Emergency Procedures for:

(1) Reactor Scram (2) Main Turbine Emergencies (3) Loss of an Instrument Bus (4) Environmental Disturbances, and (5) Loss of a DC Bus These revisions were determined to be necessary as a result of the study which showed that certain valves required manual closing to ensure availability of the firewater cooling system following a seismic event and subsequent LOFC.

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__~ ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 I.C.5 Procedures for Feedback of Operating Experience to the Plant Staff Action Requirement:

In accordance with Task Action Plan I.C.5, Procedures for Feedback of Operating Experience to Plant Staff (NUREG 0660), each applicant for an operating license shall prepare procedures to assure that operating information pertinent to plant safety originating both within and outisde 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) Identi fy 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; 9

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, ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action plan Requirements NUREG 0737 (5) Assure that plant personnel do not routinely receive extraneous and unimportant information on operating experience in such volume that is would obscure priority information or otherwise detract from overall job performance and proficiency; (6) Provide suitable checks to assure that conflicting or contradictory information is not conveyed to operators and other personnel until resolution-is reached; and, (7) Provide periodic internal audit to assure that the feedoack program functions effectively at all levels.

Procedures governing feedback of operating experience to plant staff shall be completed and the procedures put into effect on or before January 1,1981, or prior to issuance of an operating license, whichever is later.

PSC Response:

An operating assessment panel has been formulated and we are in the process of assessing the various information available for use in the operating experience feedback program. We are experiencing considerable difficulty in establishing a meaningful program in that most of the industry information addresses water reactor experience. We feel that it is extremely important that we do not burden our operators with superfluous information that is of no real value in feedback experience program while at the same time ensuring +. hat they receive necessary information. We already have a rather extensive information program that we are assessing in detail, but because of our particular unique situation we feel it will require additional time to develop and implement a good program.

In this respect we do not feel we can meet the full intent of this action requirement until January 31, 1981.

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--. ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 I.C.6 Guidance on Procedures for Verifying Correct Performance of Operating Activities Action Reouirement:

It is required (from NUREG 0660) that licensees' procedures 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 errors and-improving the quality of normal oeprations.

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.

The American Nuclear Society has prepared a draft revision to ANSI Standard N18.7-1972 (ANS 3.2)

" Administrative Controls and Quality Assurance for the Operational Phase cf Nuclear Power Plant."

A second proposed revision to Regulatory Guide 1.33,

" Quality Assurance Program Requirements (Operation), which is to be issued for public comment in the near future, will ensorse the latest draft revision to ANS 3.2 subject to the following supplemental provisions:

(1) Applicability of the guidance of Section 5.2.6 should be extended to cover surveillance testing in addition to maintenance.

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ATTACHMENT 1 Fort St. Vrain Unit No. 1 Resoonse Clarification of TMI Action Plan Requirements NUREG 0737 (2) In lieu of any designated senior reactor operator (SRO), the authority to release systems and equipment for maintenance or surveil ~.ance testing or return-to-service may be delegated to an on-shift SRO, provided provisions are made to ensure that the shift supervisor is kept fully informed of system status.

(3) Except in cases of significant radiation exposure, a second qualified person should verify correct implementation of equipment control measures such as tagging of equipment.

(4) Equipment control procedures should include assurance that control room operators are informed of changes in equipment status and the effects of such changes.

(5) For the return-to-service of equipment important to safety, a second qualified operator should verify proper systems alignment unless functional testing can be performed without compromising plant safety, and can prove that all equipment, valves, and switches involved in the activity are correctly aligned.

Licensees / applicants must review and revise procedures as necessary to reflect this position by January 1, 1981, or prior to fuel load, whichever is later.

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_.. ATTACHMENT 1 Fort St. Vrain Unit No. I Response Clarification of TMI Action Plan Requirements NUREG 0737 PSC Response:

We just received the proposed Draft 2 to ANS 3.2 referenced in the action requirements. Based on a preliminary review of this document it will be impossible for us to revise all the various plant procedures by January 1, 1981, to reflect the requirements of ANS 3.2, especially in the area of I

independent verification of equipment control measures. The major portion of our safety systems including our emergency core cooling system are in normal operation when the plant is running.

Problems with system alignment, clearance, etc., on these types of systems can be identified by the operator as a part of routine plant operation.

We are evaluating the impact of ANS 3.2, Draft 2, in more detail, but based on a literal interpretation it appears that we could not meet these requirements without increasing i

the plant operating staff. An increase to the plant staff and obtaining qualifications for the additional staffing could not be realized by January 1,1981. We will provide a more detailed response on our plans in this area by February 16, 1981.

I.D.1 Control Room Desian Reviews Action Requirement:

In accordance with Task Actiori Plan I.D.1, Control Room Design Reviews (NUREG 0660), all licensees and applicatns 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 exoected 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 significant human factors and deficiencies.

These applicants will be required to complete the more detailed control room reviews on the same schedule as licensees with operating plants.

_ ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 Operating reactors and applicants for OL's:

Complete review, using NRC guidelines (NUREG 0700) issued in 1981, on a schedule that will be determined upon issuance of the guidelines.

PSC Response:

We have NUREG/CR-1580 under review.

When NUREG 0700 is issued along with implementation schedule we will provide any further response as may be appropriate.

I.D.2 Plant Safety Parameter Display Console Action Requirement:

In accordance with Task Action Plan I.D.2, Plant Safety Parameter Display Console (NUREG 0660) 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.

Schedules for implementation will be issued in conjunction with issance of NUREG 0696.

PSC Response:

We have reviewed and commented on the preliminary draft of NUREG 0696.

In our meeting with the staff on December 10, 1980, we were informed that the NUREG has not as yet been issued. Upon isuance of NUREG 0696 we will provide further response to the requirements and schedules as may be appropriate.

II.B.1 Reactor Coolant Vents Action Reouirement:

Each applicant and licensee shall_ install reactor coolant system (RCS) and reactor vessel head high point vents

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_ _ _,.. ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 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 10CFRPart50, " 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 10CFR50.46.

(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.

PSC Resoonse:

The nuclear reactor at Fort St. Vrain is cooled by helium gas.

providing reactor high point vents on a HTGR is not necessary.

Therefore, the recommended installation of reactor coolant system and reactor vessel head high -point vents remotely operated from - the control room is not applicable to Fort St. Vrain.

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. - _ _... _ _ ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action plan Requirements NUREG 0737 The NRC Staff's evaluation of this topic, forwarded to PSC by an NRC letter received on March 31, 1980, was in concurrence with PSC's conclusion.

II.B.2 Design Review of Plant Shielding and Environmental Qualification for Spaces / Systems Which May Be Used in Post Accident Operations Action Requirement:

With the assumption of a post accident release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4 (i.e., the equivalent of 50% of the core radiciodine, 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, j

and instrument areas, in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during post accident operations of these systems.

Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, l

increased permanent or temporary shielding, or post accident l

procedural controls.

The design review shall determine which types of corrective actions are needed for vital areas tnroughout the facility.

(1) For Vital Area Access By January 1, 1982, modifications should be completed:

For operating plants, documentation should be completed by January 1, 1982.

For OL applicants, documentation of the evaluation should be completed at least four months before the operating license is issued.

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, ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 (2) For Equipment Qualification All safety-related electrical equipment must be fully qualified by June 30, 1982.

Documentation in accordance with:

(a) Operating Reactors and NTOL (operating license expected by

February, 1981):

submittal to be received no later than November 1, 1980.

(b) Operating Licenses (operating license expected by June 30, 1982):

submittal no later than 4 months before issuance of operating license.

Operating licenses in accordance with review schedule.

For Operating Reactors - By January 1, 1981, have available for review the final design details of the implementation of the above position and clarifications.

If deviations to the above position or clarification are necessary, provide detailed explanation and justifications by January 1, 1981.

For Equipment Qualification Provide the information required by the Commission Memorandum and Order on equipment qualification (CLI-80-21).

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Fort St. Vrain Unit No. 1 Response Clarification of TMI Action plan Requirements NUREG 0737 i

PSC Response:

Public Service Company has fully addressed this position in previous submittals dated December 12, 1979, (P-79299) and December 27, 1979,(P-79312). However, subsequent analysis a

l has revealed that an error in the computer model used to evaluate control room dose due to a source term in the reactor plant exhaust filters led to an incorrect conclusion regarding the need for additional shielding for those filters. The corrected analysis shows that control room doses are within GDC-19 guidelines without additional i

shielding, thus no additional reactor plant exhaust filter i

shielding is required.

The previous PSC submittal dated December 27, 1979, has been revised accordingly and is as j

follows:

PSC December 27, 1979, (P-79312) Section 2.1.6.b SUMBITTAL:

J The assessment of post accident operator actions in vital areas at Fort St.

Vrain indicates that doses received from a hypothetical FSV accident scenario will not be in excess of the GDC-19 guidelines. for the duration of the accident.

The hypothetical Fort St.

Vrain accident scenario consists of the FSV Design Basis Accident (DBA) #1 ccmbined with successive PCRV primary coolant leakage after depressurization.

For clarification, the DBA #1 and PCRV leakage scenarios are explained in Attachment 2.

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Response Clarification of TMI Action Plan Requirements NUREG 0737 4

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II.B.3 Post Accident Capability Action Requirement:

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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) 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 i

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 i

capability to promptly quantify (in-less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) certain radionuclides that are indicators of the degree of l

core damage.

Such radionuclides are noble gases (which indicate cladding failure),

iodines and cesiums (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 auxiliary building and possible contamination and direct radiation from airborne effluents.

If the review indicates that the i

analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or i

equipment procurement shall-be undertaken to meet the criteria.

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.... ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of iMI Action Plan Requirements NUREG 0737 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 baron sample analysis within an hour and the chloride sample within a shift).

Installation should take place by January 1, 1982.

Operating Reactors - By January 1, 1982, have available for review the final design details of the implementation of the above position and clarifications.

The final design includes piping and instrumentation diagrams (P&ID's),

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.

If deviations to the above position or clarification are necessary, porvide detailed explanation and justification for the deviations by January 1,1982.

PSC Response:

We responded to the action requirement in answer to recommendation 2.1.8.a of NUREG 0578.

Our reponses and additional clarification are contained in P-79299, December 12, 1979; P-79312, December 28, 1979; and P-80028, February 20, 1980.

We believe these responses adequately address the action requirement, but we are performing a more detailed review of the clarification provided by NUREG 0737.

Our new radiochemistry lab should be operational by March, 1981.

Further modifications, if any, will be implemented by Janua ry, 1982.

i

.. _ ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 II.B.4 Training for Mitigating Core Damage Action Requirement:

Licensees are required to develop a training program to teach the use of installed equipment and systems to control or mitigate accidents in which the core is severly damaged.

They must then implement the training program.

Shift technical advisors and operating personnel from the plant manager through the operations chain to the licensed operators shall receive all the training indicated in to H. R. Denton's March 28, 1980, letter.

Managers and technicians in the Instrumentation and Control (I&C), health physics and chemistry departments shall receive training commensurate with the responsibilities.

Licensees with operating reactors will develop a training program by January 1, 1981, and initiate the training program by April 1, 1981.

The initial program should be complete by October 1, 1981.

PSC Response:

Many of the training requirements outlined in Enclosure 3 to r

H. R. Denton's March 28, 1980, letter are clearly not applicable to an HTGR, and in other areas it is somewhat difficult to draw a parallel of water reactor training requirements versus HTGR training requirements. We have, however, developed our training program outline which we feel meets the intent of the action requirement and we will have training initiated by April 1, 1981.

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, ATTACHMENT 1 Fort St. Vrain Unit No.__1 Response Clarification of TMI Action Plan Requirements NUREG 0737 II.D.1 Performance Testing of BWR and PWR Reactor Safety Relief Valves Action Requirement:

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.

This requirement applies to-all operating reactors and operating license applicants.

PSC Response:

PSC has responded to this action requirement in P-80028 as a part of Item 2.1.2.

II.D.3 Indication of Relief and Safety Valve Positior/s Action Requirement:

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.

This requirement applies to all reactor licenses and applicants for operating license.

(Operating reactor licenses ccmpleted this requirements by January,.1980.)

PSC Response:

PSC responded to this action requirement as Item 2.1.3a in P-79249 and P-80028.

II.E.1.1 Auxiliary Feecwater System Evaluation Not applicable to Fort St. Vrain.

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- ATTACHMENT 1 Fort St. Vrcin Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 II.E.I.2 Auxiliary Feedwater Syster., Automatic Initiation and Flow Indication Action Requirement:

Consistent with satisfying the requirements of General Design Criterion 20 of Appendix A to 10CFR Part 50 with respect to the timely initiation of the auxiliary feedwater system ( AFWS),

the following requirements shall be implemented in the short term:

(1) The design shall provide for the automatic initiation of the AFWS.

(2) The automatic initiation signals and circuits shall be designed so that a single failure will not result in the loss of AFWS function.

(3) Testability of the initiating signals &nd circuits shall be a feature of the design.

(4) The initiating signals and circuits shall be powered from the emergency buses.

(5) Manual capability to initiate the AFWS from the control room shall be retained and shall be implemented so that a single failure in tre manual circuits will not result in the loss of system function.

(6) The ac motor-dirven pumps and valvas in the AFWS~

shall be included in the automacic actuation (simultaneous and/or sequential) of the loads.onto the emergency buses.

(7) The automatic initiating signals and circuits shall be designed so that their failure will not result in the loss of manual capability to initiate the AFWS from the control room.

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-, ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 In the long term, the automatic initiation signals and circuits shall be upgraded in accordance with safety grade requirements.

PSC Response:

In Public Service Company's response dated December 12, 1979, (P-79299), Table 2.1.7b-1 listed the feedwater flow instrumentation provided at Fort St.

Vrain.

It has subsequently been found that existing safety grade indication of feedwater (including emergency feedwater) flow was omitted.

This instrumentation,-consisting of flow monitors FM-2209, FM-2210, FM-2211, FM-2212, FM-2213, and FM-2214, is part of the Plant Protective System (PPS) and is located on'the PPS Instrument Panel (I-9310) in the Control Room.

This instrumentation is fully safety grade and satisfies the requirements of this NRC position.

A revised Table 2.1.7b-1 is submitted as Attachment 3.

As stated in an NRC letter received by PSC on March 31, 1980, the NRC staff's evaluation of the implementation o f.

these requirements concluded that Fcrt St. Vrain meets-the requirements of NUREG 0578.

In light of this, PSC's prior commitment to provide safety grade emergency feedwater flow indication in the Control Room by. January 1,

1981, is satisfied by the existing instrumentation.

II.E.3.1 Emergency Power Supply Pressurizer Heater Not applicable to Fort St. Vrain.

II.E.4.1 Dedicated Hydrogen Penetration 3

Not applicable to Fort St. Vrain.

4 ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Recuirements NUREG 0737 II.E.4.2 Containment Isolation Dependability Action Requirement:

(1) Containment isolation system designs shall comply with the recommendations of Standard Review Plan 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 give careful consideration to the definition of essential and nonessential systems, icenti fy each system determined to be essential, identify each system determined to be nonessential, describe the basis for selection of each essential system, modi fy their containment isolation designs accordingly, and report the results,* the reevaluation to the NRC.

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

.(4) The design af 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 or the Staff Interim Position of October 23, 1979, must be sealed closed as defined in SRP 6.2.4, item valves must be verifiea to be closed at least every 31 days. (A copy of the Staff Interim Position is enclosed as Attachment 1).

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ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 (7) Containment purge and vent isolation valves must close on a high radiation signal.

As part of Category "A" lessons-learned requirements, all operating plants were required to be in conformance with positions 1 through 4 by January 1, 1980.

Each licensee will provide, and justify, the minimum containment pressure that will be used to initiate containment isolation as stated position 5 by January 1, 1981. By July 1, 1981, all operating plants must be in compliance with position 5.

All operating plants must be in compliance with position 6 by January 1, 1981.

All operating plants must be in compliance with position 7 by July 1, 1981.

PSC Response:

Public Service Company has responded to position (1) through (4) in submittals dated December 12, 1979, (P-79299) and February 20,

1980, (P-80028).

NRC concluded that those positions "have been properly addressed and that Fort St.

Vrain " containment" i solation design is acceptable" in a le:ter to PSC received on March 31, 1980.

The-following information is submitted in response to positions (5), (6), and (7):

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- ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 At Fort St.

Vrain, the " containment" consists of the concrete PCRV and the interspaces between the primary and secondary closures at PCRV penetrations.

The " containment" pressure in the interspaces is always maintained above primary coolant pressure to ensure that no primary coolant helium can flow ir.to " containment" if a leak develops in the primary coolant boundary or into the environment if a leak develops ta the secondary closure.

The normal operating containment pressure is 710 psig and the normal reactor coolant pressure is about 5-15 psi lower.

If high pressure occurs in the interspace between primary and secondary closures, the normal pressure source (the purified helium head) is automatically isolated at 788 psi. The design pressure for these interspaces is 845 psi.

This design, which is unique to the HTGR, is entirely different from that envisioned in position (5), thus that position is not considered applicable to Fort St. Vrain.

The design of Fort St. Vrain does not require provisions to purge and vent any secondary containment space, thus positions (6) and (7) are likewise not applicable.

PSC believes that the applicable requirements of this section II.5.4.2 have been properly addressed in the design of the containment of Fort St.

Vrain and that the containment isolation provisions at Fort St.

Vrain do not l

require modifications.

II.F.1 Additional Accident Monitoring Instrumentation Action Requirement:

Noble gas effluent monitors shall be installed with an extended range designed to function during accident conditions as well as during normal operating conditions.

Multiple monitors are considered necessary to cover the.

ranges of interest.

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. ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 Noble gas effluent monitors with an upper range capacity of IE+5 microcuries/cc (Xe-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 (ALARA)) concentrations to a maximum of 1E+5 microcuries/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.

This requirement applies to all operating reactors and applicants for operating license.

Implementation must be completed by January 1, 1982.

Operating Reactors - By January 1, 1981, operating reactors should have available for review the final design details of the implementation of the above position and clarifications.

If deviations to the above position or clarification are necessary, provide detailed explanation and justification for the deviations by January 1, 1981.

l PSC Response:

l l

This action requirement was addressed as Item 2.1.8.b in l

P-79249 and P-79299, and it does not appear that any further action is required.

.We are, however, reviewing the clarification provided by NUREG 0737 to ensure that no further action is required.

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. --.....- ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Reouirements NUREG 0737 II.F.1, Attachment 2 Sampling and Analysis of Plant Effluents Action Requirement:

Because idoine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radiciodines for the accident condition shall be provided with sampling conducted by adsorption on charcoal or other media, followed by onsite laboratory analysis.

This requirement was originally issued by letters to all-operating power plants dated September 13,

1979, and October 30, 1979.~

This requirement was inadvertently omitted from NUREG 0660.

Significant changes in requirements or guidance are:

(1) Changes implementation date to January 1,1982.

(25 ecifies a shielding basis design envelope for ign of samplers and sample transport devices.

'ies provisions for isokinetic sampling.

(3) %

(4) Specifies representative sampling per criteria of ANSI N131-1969.

(5) Allows use of gamma radiation measurement and shielding / distance factors in lieu of analysis - of highly radioactive samples.

i This requirement will be implemented by January 1, 1982.

By January 1, 1981, operating reactors should have available for _ review the final design details of the implementation of i

l the. above position and clarifications.

If deviations to the l

above position or clarification are necessary, provide i

detailed explanation and justification for the deviations by January 1,1981.

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. ATTACHMENT 1 Fort St. Vrain Unit No. I Response Clarification of TMI Action Plan Reouirements NUREG 0737 PSC Response:

PSC responded to this action requirement as item 2.1.8.b in P-80028, P-79312, P-79299, and P-79130. We do not feel any further action is required.

We also note that Technical Specifications are required as a part of this requirement.

Based on our previous response we can see no requirement to modify the existing Technical Specifications.

II.F.1, Attachment 3 Containment High Radiation Monitor Action Requirement:

In containment radiation-level monitors with a maximum range of 1E+8 rad / hour shall be installed. A minimum of two such

'9 monitors that are physically separated shall be provided.

Monitors shall be developed and qualified to function in an accident environment.

Implementation for operating reactors must be completed by January 1, 1982.

i By July 1, 1981, have available for Operating Reactors review the final design details of the implementation of the above position and clarifications.

If deviations to the above position or clarifications are necessary, provide a detailed explanation of and justification for the deviations l

l by July 1, 1981.

PSC Response:

See response to. Item II.F.1, Attachment 2.

II.F.1 Attachments 4, 5, and 6 Not applicable to Fort St. Vrain.

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.... ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action plan Requirements NUREG 0737 II.F.2 Instrumentation for Detection of Inadequate Core Cooling Action Requirement:

Licensees 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 monitors) 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.

By January 1, 1981, the licensee shall provide a report detailing the planned instrumentation system for monitoring of ICC.

The report should contain the necessary information, either by inclusion or by reference to previous submittals including pertinent generic reports, to satisfy the requirements which follow:

(1) A description of the proposed final system including:

i (a) A final design description of additional l

instrumentation and displays; (b) A detailed description of existing instrumentation systems (e.g.,

subcooling meters and incore thermocouples), including l

parameters ranges and displays, which provide l

operating information pertinent to ICC l

considerations; and (c) A description of any planned modifications to the instrumentation systems described in item 1.b above.

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k ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 (2) The necessary design

analysis, including evaluation of various instruments to monitor water

level, and available test data to support the design described in item 1 above.

(3) A description of additional test programs to be conducted for evaluation, qualification, and calibration of additional instrumentation.

(4) An evaluation, including proposed actions, on the conformance of the ICC instrument system to this

document, including Attachment 1 and Appendix A.

Any deviations should be justified.

(5) A description of the computer functions associated with ICC monitoring and functional specifications for relevant software in the process computer and other pertinent calculators. The reliability of nonredundant computers used in the system shc;ld be addressed.

(6) A current schedule, including contingencies, for installatiLn, testing and calibration, and implementation of any proposed new instrumentation or information displays.

(7) Guidelines for use of the additional instrumentation, and analyses used to develop these procedures.

(8) A summary of key operator action instructions in the current emergency procedures for ICC and a description of how these procedures will be modified when the final monitoring system is implemented.

(9) A description and schedule commitment for any additional submittals which are needed to support the acceptability of the proposed final instrumentation system and emergency procedures for ICC.

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-- ATTACHMENT 1 Fort St. Vrain Unit No.1 Resp wise Clarification of TMI Action Plan Requirements NUREG 0737 PSC Response:

Public Service Company has reviewed this position and the latest clarifications.

PSC's submittal dated December 12, 1979 (P-79299) relative to NUREG 0578, Section 2.1.3.b, fully addresses this position and explains why PSC concluded that the existing instrumentation and controls satisfy this position.

The NRC staff's evaluation of this topic, forwarded to PSC by an NRC letter received on March 31,

1980, led to the conclusion that "the licensee does not need to provide any additional instrumentation to detect inadequate core cooling, and therefore, satisfies this requirement."

These conclusions remain valid relative to the latest clarifications and PSC considers this position satisfied.

II.F.2, Attachment 1 Design and Qualification Criteria for PWR Reactor In Core Thermocouples Not applicable to Fort St. Vrain.

II.G.1 Emeroency Power for Pressurizer Equipment Not applicable to Fort St. Vrain.

II.K.2.2 through II.K.3.29 Not applicable to Fort St. Vrain.

(Item II.K.3.17 is listed in NUREG 0737 as being applicable to all operating reactors but based on -guidance from Special Projects this item was considered inapplicable to Fort St. Vrain.)

II.K.3.30 Revised Small Break LOCA Methods to Show Compliance with 10CFR50, Aopendix K Not applicable to Fort St. Vrain.

_. ~... _ _. - - ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 II.K.3.31 Plant-Soecific Calculations to Show Compliance with 10CFR50.46 Not applicable to Fort St. Vrain.

II.K.3.44 through II.K.3.57 Not applicable to Fort St. Vrain.

III.A.2 Improving Licensee Preparedness Action Requirement:

Each nuclear facility shall upgrade its emergency plans to provide reasonable assurance that adequate protective i

measures can and will be taken in the event of a radiological emergency.

Specific criteria to meet this requirement is delineated in NUREG 0654 (FEMA-REP-1),

" Criteria for Preparation and Evaluation of Radiological Emergency Response-Plans and Preparation in Support of Nuclear Power Plants."

Schedule for Operating Reactors - For operating reactors the following implementation milestones shall be met to address the four basic elements of the introduction to Appendix 2 to NUREG 0654.

Milestones are numbered and tagged with the following code; a-date, b-activity, c-minimum acceotance criteria. They are as follows:

(1) a.

January 2, 1981 b.

Submittal of radiological emergency response plans c.

A description of the plan to include elements of NUREG 0654, Revision 1, Appendix 2

. ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Reouirements NUREG 0737 (2).

a.

March 1, 1981 b.

Submittal of implementing procedures c.

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

(3) a.

April 1, 1981 b.

Implementation of radiological -emergency response plans c.

Four elements of Appendix 2 to NUREG 0654 with 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 the existing technical specifications as the baseline or an element 1 program and/or element 2' system of Appendix 2 to NUREG 0654, or two independent element 2' systems shall provide the basic meteorological parameters (wind direction and speed and an indicator of atmospheric stability) on display in the control room.

An operable dose calculational methodology (OCM) shall be in use in the control room and at appropriate emergency response facilities.

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pgp - ATTACHMENT I Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 The following compensating actions shall be taken by the licensee for this alternative:

(1) if only element I or element 2 is in use:

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

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.

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.

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

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

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... _. _ _ ~. _. ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Recuirements NUREG 0737 (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 10CFR 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 :he 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 to NUREG 0654, with exception of the Class 8 model of element 3.

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ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 (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 a 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

X (NRC)

(e.g., 1.97 Para-meters)

Full Met. Data X

X X

(1.23 Parameters)

DCM (for Dose X

X X

X Projections)

Class A Model to X

X X

X Plume Exposure (EPZ)

Class B Model or X

X X

Class A Model (to Ingestion EPZ)

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_. _.. _ _ _ ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 (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 cc qsi stency with the operational Class A model.

Thus, actions recommended during the initial phases of a radiological emergency would be consistent with those after the TSC and EOF are activated.

(7) a.

September 1, 1982 b.

Description of the Class 8 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 Schedule for Near-Term Operating Licenses - For applicants for an operating license, at least milestones 1, 2,

and 3'

shall be met' 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

.i 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 Nuclear Power Plants," Section 2.3.3 program as the baseline or. element I and/or element 2 systems.

.. - ATTACHMENT 1 Fort St. Vrain Unit No. I Response Clarification of TMI Action Plan Requirements NUREG 0737 Complete updated emergency plans shall be provided by January 2, 1981, and complete implementing procedures shall be submitted by March 1, 1981.

PSC Response:

Our original Radiological Emergency Response Plan (RERP) was submitted on March 18, 1980, by P-80083. We had an onsite review meeting in May, 1980, and subsequently received your comments on our plan July 28, 1980. We revised our plan and resubmitted it along with our response to your comments via P-80288 dated August 28, 1980. We met with the staff on December 10,

1980, in hopes of obtaining any additional comments on the revised plan.

We learned at the December 10, 1980, meeting that our letter P-80288 had been inadvertently misplaced, and that it was never reviewed. We were informed at the staff meeting thal comments would be forwarded to us as soon as possible. To date we have not received these comments.

Based on the above outlined delays it will be impossible for us to submit a revised RERP by January 2, 1981.

It is requested therefore that we be allowed thirty (30) days to submit a revised RERP from the date we receive NRR comments.

Given the delays in the January 2,1981, date which are beyond our control we.will require additional time to l

complete our implementing procedures which are due on l

March 1, 1981. We cannot assess how much additional time might be required until we receive comments on the submitted revised RERP as we are unable to assess the extent and impact of your comments on the implementing procedures.

We do intend to meet the March 1, 1981, date for submittal of methods and programs for meeting Stage I requirements of NUREG 0654, Appendix 2, Revision 1.

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ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Reauirements NUREG 0737 We plan to meet the schedule dates of March 1, 1982, and July 1, 1982, with the reservation that these milestone dates are difficult to assess without the guidance of NUREG 0696 and the impact NUREG 0696 may have on emergency response facilities and interfaces with the meteorological requirements of NUREG 0654.

With reference to the September 1, 1982, and June 1, 1983, requirements for the Class 8 model, we are unable to assess our capability to meet these dates.

Requirements for the Class B model do not appear to be attainable given the present state of the art, and a Class B model that will fulfill the requirements does not exist. We must therefore reserve any commitment to the September 1, 1982, and June 1, 1983, dates until the requirements can be better defined and until sufficient information is available to permit an intelligent evaluation.

III.D.1.1 Integrity of Systems Outside Containment The action requirements were addressed in P-79299 as a part of Item 2.1.6.a.

No further action is anticipated.

III.D.3.3 Improved Iodine Instrumentation Under Accident Conditions Action Requirement:

(1) 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.

(2) 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 radiciodine in areas within. the plant where plet personnel may be present during an accident.

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- ~ = ~ - - - - ~ ~ ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 Licensees and applicants shall meet position 1 by January 1, 1981, or prior to licensing, whichever is later.

PSC Response:

PSC responded to this matter in P-80028 and P-79312 as a p/

part of Item 2.1.8.c.

The bases of these responses are j

still applicable to this action requirement.

With reference to the capability of removing a sampling cartridge to a low background area for further analysis we have set up an interim multi channel analyzer which is outside of the plant to obtain analyses.

Our new radiochemistry lab is presently under construction and is nearing completion, but based on delays experienced in obtaining HEPA and charcoal filters the facility will not be completely operational until March 15, 1981.

Additional iodine samplers have been obtained to meet our initial commitment and the intent of this action requirements.

Although we feel we can meet the intent of the January 1, 1981, requirements it should be noted that our permanent facilities will not be available until March 15, 1981.

III.D.3.4 Control Room Inhabitability Action Reauirement:

i In accordance with Task Action Plan item III.D.3.4 and control room habitability, 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 safely operated or shutdown under design basis accident conditions (Criterion 19, " Control Room," of Appendix A, " General Design Criteria for Nuclear Power Plants," to 10CFR Part 50).

_ _ - _ ATTACHMENT 1 Fort St. Vrain Unit No. 1 Response Clarification of TMI Action Plan Requirements NUREG 0737 Licensees shall submit their responses to this request on or before January 1, 1981. 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, if any, identified by the staff during its review will be specified to licensees.

PSC Response:

Fort St.

Vrain was reviewed against Standard Review Plans 2.2.1, 2.2.2, and 2.2.3 concerning potential

hazards, against SRP 6.4 concerning control room protection; and against Regulatory Guides 1.78 and 1.95 concerning specific guidelines for control room habitability during chemical and especially chlorine release accidents.

Although Fort St.

Vrain is not in total agreement with the guidelines, the intent of the guidelines is met. There are no areas where the degree of protection is considered to be significantly less than that specified. This is-primarily due to two factors:

1) the design of the control room ventilation system that allows isolation of the control room atmosphere, and 2) the presence of a Breathable Air System that. allows continuous occupation of the control room during all postulated accidents.

The control room habitability resulting from postulated accidental release of radioactivity and control room operator radiation exposure from plate out problem during j

design basis accidents are identified and discussed in Section II.B.2.

PSC letter P-80288 dated August 28, 1980, i

discussed,the need to supply potassium iodide on site j

(reference Attachment 4).

I

REFERENCE ITEM II.B.2 To obtain a post-accident release of radioactivity equivalent to that described in Regulatory Guides 1.3,1.4 and 1.7 requires a permanent loss of all forced circulation for the FSV HTGR. This specific accident was identified as DBA #1 in FSAR Section 14.10 and Appendix D.

These analyses performed by General Atomic Company at the time of licensing did not consider Regulatory Guides 1.3 and 1.4 source tenns (i.e., the equivalent of the 50%

of the core radioiodine and 100% of the core noble gas inventory for release to the primary coolant) appropriate for the HTGR. However, because of past precedence by the then Atomic Energy Comission (AEC) of using the above source terms, offsite doses resulting from the postulated accident were calculated and presented in the previously mentioned FSAR sections using both the General Atomic Company release assumptions and AEC TID-14844 release assumptions.

In both cases the offsite doses are within 10CFR100 limits.

DBA #1

Description:

. A non-mechanistic loss of forced circulation is postulated from full power operation, where the reactor is scrammed by the plant protective system and all attempts to restore forced circulation using the multiple heat sinks, circulators and motive power for the circulators fail. Because of the large heat sink provided by the graphite core, considerable time is available to initiate primary coolant depressurization and to restore forced circulation.

The FSV FSAR specifies the time available to initiate depressurization to be 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, which was later amended by PSC letter P-77250 dated December 22, 1977 to be 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The reduction in time was due to the capability of the helium purification system to process primary coolant during the planned blowdown of the clean primary coolant to the reactor building ventilation stack. Thus, the depressurization of the PCRV is initiated after 2 hou,s and completed 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> later (or 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> from the onset of the accident), at which time the PCRV has been depressurized to 5 psig.

The fuel is slow to heat up due to the large heat sink provided by the core graphite. A peak average active core temperature of 5400 degrees F is reached about 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> after the onset of the accident. At this temperature, the core structural integrity and geometry are not compromised since the vaporization temperature of graphite is 6900 degrees F.

Peak activity released to the primary coolant, considering decay, is reached about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into the accident.

Heat removal is provided by the liner cooling system in the redistribute mode which maximizes cooling in the top head of the PCRV.

Leakage of primary coolant from the PCRV is assumed to occur at a conservatively high leakage rate of 0.2% of the primary coolant inventory per day.

Offsite doses were calculated for a 6 month duration of the accident, but most of the offsite dose occurs in the first 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> of the accident, due to fission product decay.

The reactor building ventilation system maintains continous venting of the reactor building environment at 1.5 volumes /hr during the entire period of the accident.

REFERENCE ITEM II.B.2 Primary Coolant Leakage Rate During DBA #1:

The FSV FSAR OBA #1 (Appendix D, page D.1-56) assumed an arbitrarily conservative and non-mechanistic estimate of PCRV leakage after the intentional depressurization by assuming that the liner has failed completely (or does not exist) and only concrete permeability controls the leakage. An internal rate of 8.33 x 10-g5 psi pressure dif#erential was assumed which purportedly gave a PCRV fraction per hr (0.2%/ day).

Reference was made to Question IX.7 of Amendment No. 2 and Question D.2 of Amendment No. 9 of the FSV PSAR for the calculation of the pemeation rate for the FSV PCRV concrete under these conditions.

Examination of Question 0.2 revealed simply the conclusion that a 5 psi positive differential pressure led to 0.2%/ day and 2 psi positive differential pressure led to 0.08%/ day. Question IX.7 also did not provide details of the calculation of the 0.2%/ day rate.

However, considerable detail and a derivation was provided for the analysis of leakage rate testi at high pressures. The following equation was provided (egn.14 on page IX.7-8):

W (lb/ day) = 1.13 x 10-[p A j g

fp]2_

+ 2.2 x 10 p2 2

(eqn.14)

Where 6P PCRV inside pressure in psig

=

6Po PCRV inside pressure in psig for which the net

=

compressive stress in concrete = 0 2

A Face area of concrete, ft

=

X Concrete thickness, ft

=

P) = Pemeation or high side pressure, psia P2=

Ambient or low side pressure, psia l

Numerical values were inserted for P)the following equation (egn.15 on same

= 845 psig with the assumption that

.6Po was approximately equal to P) in page):

W = 1.13 x 10-5 9 0 5 + 9.1 x 10-7 9 0 2

2 x

in 857.5 12.5 l

= 0.043 + 602 = 600 lb/ day (egn.15) l The first item to note is that the coefficient for the second (laminar l

flow) tem is in error which is most likely a single error 1n transcribing l

from equation 14 to 15 since equation 13 has the 9.1 x 10-7 coefficient.

l Equation 15 should read:

W = 1.13 x 10-5 9000 8

.5

+ 2.2 x 10-6 9 0 2

2 x

in 857.5 12.5

= 0.043 + 1445 = 1450 lb/ day (eqn.15 l

revised)

Attachment'2 REFERENCE fTEM II.B.2 The second item is that the LP/1.P., term has been dropped in going from egn.14 to egn.15, which is significant if it is assumed that these equations are appropriate for evaluating the leak rate at Pj = 5 psig.

LEAK RATE Pressure P lb/ day

%/ day j

(psig)

Eqn 14 15 15 Revised 14 15 15 revised Given App D; Amend 9 Question D.2 5

.0019

.13

.30

!.0 01

.07

.17

.20-1 i

Amend 9 Question D.2 2

.0003

.046

.107

.0001

.025

.059

.08 Since equation 14 is the appropriate equation, the 0.2%/ day leak rate is conservative by a factor of 200.

Furthermore, the only equation that comes close to the values given in the SAR is 15 Revised, that is, L P/'.P, has been neglected which accounts for the factor of 200.

For purposes of plant shielding and equipment environmental evaluations, the historic 0.2%/ day is assumed to exist as an upper limit of all potential contaminated primary coolant leakage including permeability through the PCRV concrete.

This is jucged to be conservative since the primary coolant with any significant activity is contained within the PCRV or helium purification components contained in wells within the PCRV.

Radionuclide Source Terms for 08A-1:

As previously stated, the fuel within the graphite core is slow to heatup during DBA#1. Once it has reached the FSAR fuel particle coating failure temperature of 1725'C (3137'F), the fission products are assumed, for purposes of this shielding evaluation, to be realeased per the TID-14844 assumptions.

For release to the primary coolant within the PCRV, this is 100% of noble gases, 50% of the iodines and 1% others.

The total activity in curies contained in the primary coolant, assuming no leakage l

from the PCRV, as a function of lapsed time, is given in Table 2.1.6.b-1.

Consistent with TIC-14844 release assumptions, 50% of the iodines plateout within the primary coolant system resulting in a depletion of the iodine to 25% of core inventory in the reactor building air.

Thus, the total activity in curies in the reactor building, assuming the upper limit of 0.2%/ day leakage (which is being purged by the reactor building ventilation system at the rate 1.5 volumes /hr), is given in Table 2.1.6.b-2.

1ABIE 2.1.6.li-1 EV-NUREG-0578 liTUDY TOTAI ACTIVITY (C1) PitESENT IN 'lllE PCRV Pit 11tARY COGI. ANT AT 01VEN EI.APSED TIHE (houre).

FCRV PHESS HOU110ARY ltEHAINS IllTACT.

TID-14844 HORHAI.1ZAT10H FRACTIONS USEli, 100% Honi.E CASES, 50Z 100lHE 1% OTill.RS ELAPSED TIME (llours) liuCI.10E 2

11 24 34 40 4 11 52 5 11 72 120 240 475 720 4320 K r-fl0 1.05104 2.89805 2.110105 2.39104 5.89103 1.37103 5.50102 1.76:02 7.04101 0

0 0

0 0

Rt.-Ill!

8.57103 2.79105 2.80105 2.66104 6.51103 1.46103 6.07102 1.89102 7.00101 0

0 0

0 0

Zr-95 3.15101 6.66103 1.84105 2.57105 3.01105 3.59 05 3.69105 3.114805 4.111805 4.12105 3.88105 3.43805 3.02105 4.60804 lih-95 3.18:01 6.74103 1.117405 2.63105 3.09:05 3.69105 3.80105 3.97105 4.351,05 4.37105 4.31:05 4.12105 3.88105 8.90804 1-131 1.33103 3.50105 6.111106 6.91106 7.33106'7.88106 7.90106 7.93806 7.98106 7.57106 4.89106 2.07106 8.45105 0

I-132 1.44803 2.34105 1.79106 6.09:05 5.64105 5.61105 3.68105 2.96105 2.72 05 1.76105 4.02104 4.99103 5.46102 0

El,g I-133 2.411:03 5.30105 6.44806 5.25106 4.70106 4.12806 3.65106 3.05 06 2.04:06 4.114:05 8.111103 0

0 0

g r

Xe-133 5.25103 1.40106 2.50107 2.7fil07 2.94107 3.14107 3.14:07 3.12:07 3.09107 2.73107 1.41107 3.86106 9.90805 0Q!

p 1-135 1.911103 2.46105 1.40806 5.49105 3.31105 1.811105 1.25105 6.113104 1.711:04 2.94102 0

0 0

0="

Xc-135H 7.211102 11.34104 4.59605 1.72105 1.04105 5.97104 3.91104 2.14:04 5.58103 0

0 0

0 0

Xc-135 1.75103 5.43105 6.24:06 3.116106 2.93106 2.11406 1.62'606 1.08106 4.39805 1.81104 0'

O O

O nn-140 5.44101 1.44104 2.58105 2.92805 3.13405 3.39805 3.42405 3.45105 3.54105 3.57505 2.70105 1.56105 8.80104 0

f.n-140 3.34101 7.37103 2.01105 2.60105 2.93105 3.36s05 3.43105 3.54105 3.75105 3.96105 3.10:05 1.80805 1.01105 0

]

I Allt.E 2.1.6.l>-2

'SV-IlulLEC-0578 STillW TOTAL. ACTIVITY (C1) PRESENT IN 'lllE HEACTOR BU11.DIHC ATil0SI'llERE AT CIVEN EI.APSI'll T l'CIW I.EAK RATE TO hull.D1HC 0.22/ DAY.

REACTOR llUII.0INC VEH'IEI) AT 1.5 Vol.UtlES/HR. T10-14844 HolulAI.IZE0 FRACTIONS USEll, 200% Nollt.E CASES, 25% 100lHE,1% 0111ERS El Al' SED 1IME (llours) riuCI.lDE 2

11 24 34 40 4 11 52 5 11 72 120 240 475 720 4120 Kr-88 3.77-01 1.31801 1.33101 1.32'100 3.22-01 7.10-02 3.00-02 9.23-03 3.30-03 0

0 0

0 0

111, - 1111 3.55-01.1.37801 1.42101 1.48100 3.58-01 7.77-02 3.34-02 1.02-0. 3 61-03 0

0 0

0 Zr-95 1.20-03 3.29-01 9.111 00 1.40801 1.64101 1.97-101 2.04101 2.12101 2.31-101 2.29101 2.16801 1.91101 1.681 tili-95 1.21-03 3.33-01 9.911100 1.43801 1.69101 2.02101 2.10101 2.19:01 2.41101 2.43101 2.39801 2.29801 2.16:01 4.9480c 1-131 2.52-02 11.64100 1.65102 1.90102 2.02802 2.17102 2.19102 2.20:02 2.21102 2.10:02 1.36102 5.76801 2.35101 0

1-132 2.57-02 5.24 00 4.24801 1.58101 1.46101 1.46-101 1.05101 8.46100 7.75800 5.14100 1.30 00 1.61-01 1.77-020 1-133 4.68-02 1.30101 1.70102 1.44102 1.29802 1.13102 1.01302 8.45001 5.65101 1.34101 2.45-01 mhg 0

0 0

"O xe-133 1.9)-01 6.94801 1.34803 1.54103 1.63803 1.75103 1.75103 1.75103 1.75003 1.52103 7.85102 2.14102 5.50801 hy 0

1-135 3.68-02 5.119800 3.60101 1.51101 9.05100 5.09100 3.47100 1.119:00 4.89-01 m

[N 0

0 0

0 0

Xa-135H 3.14-02 7.71800 8.59301 7.311101 5.53101 3.53801 2.75101 1.81101 6.20100 1.07-01 9

0 0

0 0

Xe-135 6.75-02 2.73101 3.40102 2.26102 1.72102 1.22102 9.56101 6.40 01 2.56101 1.01800 0

0 0

0 lin-140 2.06-03 7.12-01 1.38101 1.61801 1.72101 1.117:01 1.119101 1.91801 1.96 01 1.98101 1.50101 8.67100 4.89100 0

1.a-140 1.27-03 3.66-01 1.08101 1.43101 1.61101 1.85001 1.90101 1.96101 2.08101 2.20101 1.72101 9.98100 5.62100' O

O

Attach: ant 2 REFERENCE ITEM II.B.2 Radiation Levels During DBA-1:

Based upon TID-14844 source term release assumptions, the radiation levels were calculated in the reactor building and the control room to detemine the operator accessibility. Details are described herein.

Assumptions In addition to the assumptions used in deriving the source tems, the following assumptions were made for evaluating shielding adequacy:

1.

Credit was taken for a 50". depletion of the iodines due to plateout in the primary coolant system prior to release to the reactor building atmosphere.

2.

All fission products were assumed to remain gasborne.

In other words, no plateout of fission products was contemplated.

3.

All the activities were uniformly distributed throughout the free space of the reactor building or the PCRV.

4.

The iodines and particulates removed by the reactor building ventilation filters were deposited in any two of the three filters available.

5.

^-lu major shielding such as concrete walls was considered.

Reactor Building To detemine the accessibility of the reactor building during the course of DBA-1, the gamma dose rate in the reactor building was calculated as a function of elapsed time. The contributing sources consist of the gasborne activity in the reactor building as a result of PCRV leakage, the primary coolant activity contained in the PCRV and the buildup of iodines and particulates in the reactor plant exhaust filters. However the contribution from the reactor plant exhaust filters was not included in this part of the analysis because the accumulation of iodines and particulates in the time frame of interest, the first 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />, is not significant.

Two dose points were selected for the dose rate calculation. The first point is located at the center of the space above the refueling floor

(=40 ft from the floor), and the second point is on the refueling floor directly above the refueling penetration.

The PATH code described in FSAR Section 11.2.2.4 was utilized to perform the calculation.

Figure 1 shows the dose rate at the first dose point.

Essentially all the contributions come from the gasborne activity in the reactor building.

The activity in the PCRV is relatively insignificant to the first dose point, because of a large separation distance between the source and dose point.

Short-tem access to the reactor building is possible.

l

REFERENCE ITEM II.B.2 The dose rate of the second dose point (i.e., the refueling floor) is given in Figure 2.

The contributions from the reactor building and from the PCRV are individually represented, along with the total dose rate. The contribution from the PCRV is due to the primary coolant activity present in the interspace below the primary closure for the control rod drive. The maximum dose rate on the floor is 1.0 rem /hr, which is less than the peak dose rate of 1.4 rem /hr at the first dose point.

Therefore, the refueling floor is accessible on a short-tenn basis.

Control Room The dose rates in the control room include the contributions from the airborne activity in the reactor building atmosphere, and from the iodine and particulate activity accumulated in the plant exhaust filters.

The PATH code was used to determine the contribution from each source as a function of time into accident.

For each source, the dose rates were calculated at various points in order to locate the point of the maximum dose rate.

The results of the PATH calculations are presented in Figure 3 as a function of elapsed time.

The contributions from the reactor building and from the exhaust filters are shown separately at different dose points.

Both curves in Figure 3 represent the maximum dose rates with respect to locations.

The location of the maximum dose rate from each source is indicated in Figure 3.

It appears that the dose rate due to the reactor building airborn<a activity is acceptable for occupational access. The filter activity gives a peak dose rate of 24 mrem /hr near the ceiling of the Reactor Engireer's Office.

The dose rate drops by more than one order of magnitude at points within the personnel level (i.e., within 6-foot height off the control roon floor).

In other words, the peak dose rate from the filter activity should be less than 2.5 mrem /hr at the personnel level.

Summary Results The peak dose rates in the reactor building and cor. trol room are summarized below. Also indicated are the time at which the peak dose rate occurs following an accident and the total dose accumulated over a period of 180 days from initiation of the accident. The 180-day dose is given for tne control room only based on the occupancy factor provided in NRC Standard Review Plan 6.4.

l Peak Gamma Time of 180-Day l

Location and Condition Dose Rate Peak 00se (rem)

Reactor Building (above refueling floor) 1.4 R/hr ID 1

Control Room (at ceiling) i From Exhaust Filters 24 mR/hr

~30D 25 From Reactor Building 2 mR/hr ID

~ 0. 3 Control Room (at personnel level)

From Exhaust Filters

<2.5 mR/hr

~ 30D

< 2. 5 From Reactor Building 3 mR/hr ID 0.4

REFERENCE ITEM ZI.B.2 Conclusions The following conclusions are recched from-the review of shielding design adequacy for DBA-1 conditions and TID-14844 source term release assumptions:

1.

The radiation levels in the control room are acceptable for occupational occupancy during DBA-1.

Both dose rate and integrated dose within the spaces occupied by the personnel meet the NRC regulatory criteria.

2.

No additional shielding is required for the plant exhaust filters, as far as personnel access to the reactor building during the first 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> of the accident and continuous access to the control room is concerned.

3.

Areas immediately outside the reactor building should be accessible only on a restricted basis because of direct raciation from the activity in the reactor building.

Attachmen' 2 q- -

. g (y

n t

- ),

e REFERENCE ITEM II.B.2 e

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REFERENCE ITEM IZ.B.2 Early Operator Actions Required Durino 08A-1:

The time frame of interest is the first 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> of the accident.

In the first two hours, operator access to the reactor building is required to connect the high temperature filter absorber cooling coil to the reactor plant cooling water system and to open the 1" vent line to the reactor building ventilation stack. The 1" valve i.; V-23279 shown in Figure 4.

At this point there has been no significant fuel failure and thus no release to the reactor building environment.

All other required operator actions are accomplished remotely from the control room. These include operation of the reserve shutdown system, placing the liner cooling system in the redistribute mode, and increasing the liner cooling water pressure by increasing the cover gas pressure from the normal 2 psig to 30 psig.

Between 3-l/4 and 3-3/4 hours into the accident, operator access to the reactor building is required to open the 2 inch valve (V-03271 in Figure 4) in the primary coolant flow path to the reactor building vent system.

About 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> from the start of the accident, operator access is again required to close the two valves (V-23279 and V-23271) at the completion of the PCRV depressurization to 5 psig.

It is estimated the transient time and time to operate the valves will require at most 5 minutes in the reactor building per event.

Pnst Accident Operator Actions Required Following DBA-1:

There are no operator actions for maintenance of core cooling which require access to the reactor building after 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> from the onset of the accident.

Following are evaluations of plant systems regarding required post accident operation capability from the control room.

1.

System 46, Reactor Plant Cooling Water System This system is operating and must continue to operate for maintenance i

of liner cooling water.

To remain operable, the following equipment must remain operable or the 4

following actions must be taken:

a)

Cooling water pumps, one per loop, must operate.

Since one pump per loop is on at the time of the accident and since they are operated remote-manual (R-M) from the Control Room (CR), this requirement is satisfied.

i b)

The operator must be capable of increasing the cover gas pressure to the surge tanks to maintain subcooling allowance.

This is R-M from the CR.

c)

The operator must maintain water level in the surge tanks.

The tanks are level alarmed and recorded in the CR.

Condensate is added R-M from the CR.

e

REFERENCE ITEM II.B.2 d)

The operator should know if there is flow in the system. There is a flow alarm and automatic standby pump start capability in the j

CR.

e)

The operator should be able to redistribute flow to the. top head and upper core barrel regions to prevent localized boiling. This is done R-M from the CR.

f)

The operator must maintain service water flow to the cooling water heat exchangers. The cooling water temperature is recorded and service water flow controlled in the CR.

2.

System 21, Helium Circulator Auxiliary System This system has been shutdown and it does not have to operate.

In the non-operating mode, it is expected that a)

The auxiliary system is shutdown b)

The circulator isolation valves are closed c)

The brake and seal are set d)

Should the seal malfunction, the static seal-backup isolation system can be employed to establish a water seal in the circulator J

and prevent leakage of radioactive helium. This is accomplished i

R-M from the CR.

3.

System 22, Secondary Coolant System This system has been shutdown and it does not have to be operated.

In the non-operating mode, it is expected that a)

The feedwater valves are closed b)

The valves to desuperheaters, preflash tanks, and main condenser l

are open so that pressure cannot build up in the steam generators c)

The steam generators eventually boil dry, at which time the valves in 3.b) above are closed R-M from the CR.

l 4.

System 47, Purification Coolino Water System l

l This system provided cooling to the high temperature filter absorber (HTFA) cooling coils during the initial PCRV depressurization.

It was then shutdown and does not have to operate in any further post accident l

operations.

l l

o REFERENCE ITEM II.B.2 5.

System 45, Fire Protection System The Fire Protection System in the Fort St. Vrain plant design serves as a backup source of cooling water to the PCRV liner cooling system.

In the unanticipated event that either of the two loops of the reactor plant cooling water rystem are not operable, fire water would be used to perfonn the PCRV liner cooling function. Local manual valve operation in the reactor building is required to use this mode of liner cooling.

Operator access to these valves is possible at any time during the course of DBA-1 since the accumulated dosage would be within GDC-l?

limits.

6.

System 48, Alternate Cooling Method The Alternate Cooling Method (ACM) provides an alternate means of providing electrical power and control for cooling the reactor in the event of the occurrence of disruptive faults or events, such as a major fire in congested cable areas or the Three Room Control Complex. The system is provided to ensure public health and safety consequences, analyzed and presented in DBA-1 in the FSAR, are not exceeded.

This system represents additional plant capability to maintain reactor cooling under DBA-1 conditions for the events stated above.

It would not be used when normal control functions are maintained from the Three Room Control Complex.

Should the ACti be required, its use would be icitiated within the first two hours of the accident.

Specifically, the following equipment in the reactor builJing would require local manual operation of the ACM transfer swite.es.

a)

PCRV Cooling Water Pump 1 A-(P-4601) b)

PCRV Cooling Water Pump 18-(P-46015) c)

PCRV Cooling Water Pump IC-(P-4602) d)

PCRV Cooling Water Pump 10-(P-46025) e)

Purification Cooling Water Pump (P-4701) f)

Purification Cooling Water Pump (P-4702) g)

Valve HV-2301 h)

Valye HV-2302 Adecuacy of Equipment and Instrumentation for DBA-1 Radiation Levels:

The 400 rem accumulated gamma dose in the reactor building for the 180 day duration of DBA-1 pose no hazard to instrumentation or equipment contained therein.

The same is true for 2700 rem accumulated dose in the vicinity of the reactor building ventilation filters.

= -

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REFERENCE ITEM II.E.1.2 TABLE 2.1.7b-1 FEEDWATER FLOW INSTRUMENTATION I. Loop 1 Feedwater/ Emergency Feedwater/ Emergency Condensate / Firewater-Flow to S/G EES Section hpe.

Number Location

• Safety Grade Flow Element FE 2205 Local Yes Flow Transmitter, High Rg.

FT 2205 Local Yes Flow Monitors FM 2205-1,2,7 I35 (AEE:!)

Yes Flow Controller FC 2205 105 CR)

Yes Flow Monitors FM 2205-4,5,6 I35 AEER)

No Flow Indicator FI 2205 I49 480V)

No Flow Indicator FI 2205-1 105(CR)

No Flow Indicator FI 2205-2 105 (CR)

No Flow Recorder FR 2205 105 (CR)

No Flow Switch Low FSL 2205-1 I70 AEER)

No Flow Alann Low FAL 2205 105 CR)

No Valve Position Indicator ZI 2205 105 CR)

Yes Flow Transmitter, Low Range FT 2207 155(TB,MEZ)

Yes Flow Monitor FM 2207 105(CR)

No Flow Controller FC 2207 135A (AEER)

No Flow Transmittcr (to PPS)

FT 2209 155 TB,MEZ)

Yes Flow Monitor / Indicator (PPS)

FM 2209 110 CR)

Yes Flow Monitor FM 2209-1 I39 AEER)

Yes Flow Transmitter (to PPS)

FT 2211 155 TB,MEZ)

Yes Flow Monitor / Indicator (PPS)

FM 2211 Il0 CR)

Yes Flow Monitor FM 2211-1 I40 AEER)

Yes Flow Transmitter (to PPS)

FT 2213 155(TB.MEZ)

Yes Flow Monitor / Indicator (PPS)

FM 2213 Il0(CR)

Yes Flow Monitor FM 2213-1 I43(AEER)

Yes II. Loop 1 Emergency Condensate / Firewater Flow to S/G. Reheater Section hge.

Number Location

• Safety Grade Flow Element FE 2293 Local Yes Flow Transmitter FT 2239 I128 (RB,EL4759')

Yes Flow Monitor FM 2239 1358(AEER)

Yes Flow Controller FC 2239 I05 (CR)

Yes Flow Recorder FR 2239 IOS (CR)

Yes III. Loop 2 Feedwater/ Emergency Feedwater/ Emergency Condensate / Firewater Flow to S/G EES Section

hpe, Number Location

• Safety Grade Flow Element FE 2206 Local Yes Flow Transmitter, High Rg.

FT 2206 154 (TB, MEZ)

Yes Flow Monitors FM 2206-1,2,7 136A (AEER)

Yes Flow Controller FC 2206 105(CR)

Yes Flow Monitors FM 2206-4,5,6 136A (AEER)

No Flow Indicator FI 2206 I49(480V)

No Flow Indicator FI 2206-1 105(CR)

No Flow Indicator FI 2206-2 105 (CR)

No Flow Recorder FR 2206 105 (CR)

No l

Flow Switch Low FSL 2205-1 170 (AEER)

No L

Flow Alam Low FAL 2206 105(CR)

No

[

Valve Position Indicator ZI 2206 105(CR)

Yes

p (REFERENCEITEMVI.E.1.2)

TABLE 2.1.7b-1 (Continued)

III. Loop 2 Feedwater/ Emergency Feedwater/ Emergency Condensate / Firewater Flow to S/G EES Section (Continued)

hpe, Number location

• Safety Grade Flow Transmitter, Low Rg.

FT 2208 154(TB,MEZ)

Yes Flow Monitor FM 2208 I36A (AEER)

No Flow Controller FC 2208 105 (CR)

No Flow Transmitter (to PPS)

FT 2210 Local Yes Flow Monitor / Indicator (PPS)

FM 2210 Il0 (CR)

Yes Flow Monitor FM 2210-1 139 (AEER)

Yes Flow Transmitter (to PPS)

FT 2212 154 (TB, MEZ)

Yes Flow Monitor / Indicator (PPS)

FM 2212 Il0 CR)

Yes Flow Monitor FM 2212-1 I40 AMR)

Yes Flow Transmitter (to PPS)

FT 2214 I54 TB,MEZ)

Yes Flow Monitor / Indicator (PPS)

FM 2214 110(CR)

Yes Flow Monitor FM 2214-1 143 (AEER)

Yes IV. Loop 2 Emergency Condensate / Firewater Flow to S/G Reheater Sections Type Number Location

• Safety Grade Flow Element FE 2240 Local Yes Flow Transmitter FT 2240 1139 (RB, EL4759')

Yes Flow Monitor FM 2240 I36B (AEER)

Yes Flow Controller FC 2240 105 CR Yes Flow Recorder FR 2240 105 CR Yes V. Emergency Feedwater Flow to Feedwater Header Type Number Location

• Safety Grade Flow Element FE 2297 Local Yes Flow Transmitter FT 2297 171 (TB, Grade)

Yes Flow Monitor FM 2297 I45(AEER)

No Flow Indicating Switches High FISH 2297,8,9 Local Yes Flow Indicator FI 2297 102(CR)

No VI. Steam Generator Module Feedwater Flow

hpe, Number Location

• Safety Grade Flow Elements FE 2222 12 Local Yes Flow Transmitters FT 2222 12 1125, 1131, 1134, 1140 (RB,EL4799')

Yes Flow Monitors FM 2222 12 1358, 1368 (AEER)

No Flow Recorder, Multipoint FR 2222 113 (CR)

No

• Location by equipment rack number and physical location.

Physical locations given as:

CR -- Control Room AEER -- Auxiliary Electrical Equipment Room 480V -- 480V Switchbear Roomilding Mezzanine Level TB, MEZ -- Turbine TB, Grade -- Turbine Building, Grade Level RS, EL4759' -- Reactor Building, Elevation 4759'

p ATTACHMENT 4 REFERENCE ITEM III.D.3.4 The control room emergency zone at Fort St. Vrain consists of the control room and adjacent areas such as the kitchen and washroom.

The control room under normal conditions is staffed with two (2) people at all times. Various other people are entering and leaving as required for plant operations.

During an emergency, the Personnel Emergency Response Plan (Administrative Procedure G-5) requires five (5) people to be in the control room full time.

Additionally, a health physics person is assigned to the control room while four (4) other Operations Department Personnel will be in and out as required.

During an emergency, which requires the personnel to use the Breathable Air System, the control room occupancy is limited to six (6) to match the number of air supply connections. Other personnel will be requested to use Scott Air Pacs if they are in the control room.

In addition to the protection offered by self-contained Scott Air Pacs and the breathable Air System, the control room ventilation system utilizes the Control Room Makeup Ventilation Filter (F-7502) which is of the CBR type and is rated at 1500 cfm but operates at a flow of 480 cfm. The filter consists of a particulate filter in series with a gas absorber containing activated charcoal. The filter is designed to meet all the requirements of the U.S. Army Chemical Corps Specification MIL-F-50052.

In addition, Control Room Ventilation Filter (F-7503) has a filter efficiency of 45% by the NBS atmosphere dust spot test, and is rated at 21,160 cfm. The filter is equipped with an upstream prefilter tc trap large particles.

In conjunction with filters F-7502 and F-7503, Control Room Charcoal Filter (F-7504) has elements with a nominal 1" thickness and is rated at 21,160 cfm. Particulate matter is removed by ventilation filter (F-7503) before passing through the charcoal.

Figure 1 (attached) is a schematic of the Control Room Ventilation System.

The following is a detailed comparison of existing FSV conditions to Standard Review Plans 2.2.1, 2.2.2, 2.2.3 and 6.4:

A.

SRP's 2.2.1-2.2.2 and 2.2.3 - Hazard Identification 1.

Guidelines I

These SRP's address the identification of potential hazards and accidents within 5 miles of the plant.

i l

l

)

ATTACHMENT 4 REFERENCE ITEM III.D.3.4 2.

FSV Existing Condition On site: Chlorine is stored in liquid form in 1-ton bottles outside the Chemical Building, about 360 feet from the Control Room. Also, various chemicals such as 29% concentrated ammonia, 93% sulfuric acid, and 50% caustic are stored in the demineralizer room on the ground floor of the turbine building. There is an indoor turbine lube oil storage tank and outdoor underground storage tanks for gasoline, diesel fuel, and No. 2 fuel oil; these could produce hazardous combustion products if they were ignited.

In addition, there is Halon and CO2 for the fire protection systa.ns.

Within 5 miles: There is a Union Pacific RR track about 3 miles east that is the main north-south line between Denver, Colorado and Cheyenne, Wyoming; it carries LP Gas and occasionally liquid chlorine.

Another tract 3/4 mil West of FSV carries mostly coal. Al so,

there are two oil lines, one 3.1 miles and one 4.7 miles from FSV, and a 4" to 6" medium pressure (140-150 psi) natural gas transmission line about 3/4 mile south of FSV. There are numerous anhydrous ammonia tanks used for fertilizer storage on adjacent farms, but there are no industrial activities that use chemicals or toxic materials.

3.

Comments a.

As will be discussed with the specific guidelines, chlorine storage and the proximity to the railroad tracks are in accordance with Regulatory Guide 1.95.

l

3 ATTACHMENT 4 REFERENCE ITEM III.D.3.4 b.

Chemicals are properly stored at the Plant inside closed systems in a room with outside vents two floors below the Control Room. The Control Room ventilation intake is 60' above the elevation of the demineralizer (chemical storage) room vent; with ammonia's toxicity level of 100 ppm and its acrid smell that is detectable at a much lower level, it is concluded that the intake dampers could be closed and the respirators donned before personnel injury.

c.

The fire protection systems are designed to minimize fires in the petroleum tanks and to alert personnel so that breathing apparatus can be used, if necessary.

d.

The oil and gas line hazards due to explosions and fires are far enough away that there can be adequate warning to control ventilation as required.

B.

SRP 5.4 - Control Room Habitability 1.

Breathing Apparatus a.

Guideline Paragraph 6.4.II.4 states that self-contained breathing apparatus for an emergency team (at least 5 men) should be on hand in the Control Room. Also, a six-hour on site bottled air supply, 30 man-hours, should be available with unlimited off-site replenishment capability from nearby locations.

b.

Existing FSV Condition There are 6 Scott Air Pacs in or immediately outside the Control Room, with 12 spare air bottles. There is also a Breathable Air system with 2 independent compressors and purifiers, each of which can provide 20 scfm to 6 masks in the Control Room. This system will remove ch'.orine and other noxious gases. There is also a 1140 scf 2400 psig storage volume that can recharge 2 Scott Air Pacs and supply 5 respirators I

for 45 minutes without recharging. This is about 24 man-hours of available air, in addition te which there is about 10 man-(

hours of air in reserve air pacs located in the rest of the plant.

It is noted here that the Breathable Air System compressors have a suction point about 8' above grade.

This keeps out dust and minim!zes the amount of heavy, dense gases (like chlorine) that get drawn in.

The filter canister 3 of the Breathable Air Compressors are rated for 40,000 f t of air, minimum. At the normal 20 CFM, each set of canisters could filter for at least 33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> and in a dry environment, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> could be expected.

FSV monitors j

the compressor elapsed time meters to insure that there is l

sufficient remaining capacity to handle accidents and replacement I

cartridges are available locally.

1

3 7

ATTACHMENT 4 REFERENCE ITEM III.D.3.4 2.

Emergency Team Support J

\\

a.

Guideline l

Paragraph 6.4.II.2 states that food, water and medical supplies should be sufficient to maintain the emergency team for 5 days.

b.

Existing FSV Conditijg1 FSV does not have this material stored.

For the FSV facility all analyzed accidents are of short duration so that these supplies will not be required.

In the event of a long term accident, these materials are available nearby and could be obtained as required.

C.

Regulatory Guides 1.95 and 1.78 are referenced by SRP 6.4 and they provide the following:

1.

Regulatory Guide 1.95 postulates two basic chlorine accident types:

a long-term low-leakage-rate release, or a short-term puff release.

For the first type, only breathing apparatus is necessary to protect the control room operator, if he is given warning.

For the second type, the control room should be automatically isolated.

For the low-leakage-rate accident, FSV has adequate breathing apparatus (see B.1 above) and is installing a chlorine leak detector at the chlorine storage facility. Although this leak detector is inside the builidng while the chlorine storage bottles are outside, most of the connections are inside so most slow leaks will be detected.

The puff release would most likely occur during loading and unloading the cylinders, which occurs about 360 feet from the Control Room ventilation intakes.

There is not a direct path between the chlorine bottle storage area and the Control Room intake, chlorine gas is heavy and would have to rise 75' to the Control Room air intakes, and significant diffusion would take place over this distance.

For these reasons, the puff release is not considered to be a significant fontrol Room hazard.

FSV meets the guidelines as discussed below.

2.

Material Storage a.

Guidelines Liquified chlorine should not be stored within 100 meters of a control room or its fresh air inlets.

Also, the largest container should have an inventory of 2000 lbs, and there should be a capability for manual isolation of the ventilation system.

For large quantities as would be in RR tankers, they should be over 2000 meters (6560') away.

Specific criteria is not provided for other substances.

a.

ATTACHMENT 4 REFERENCE ITEM III.D.3.4 It is noted that the design concept for the Control Room ventilation system includes a full flow activated carbon filter that is nonnally bypassed but would be put on line when the inlet dampers are closed. Also, the makeup filter includes an activated carbon section that is rated for chemical, biological and radiological service. The full flow carbon filter has the cagacity to absorb about 20 pounds of chlorine or about 100 ft of pure chlorine gas at STP.

Since the filter is not placed in service until after an accident, this capacity is considered adequate for cleanup of.the initial concentrations of chlorine in the control room that entered before the area could be isolated.

4.

Breathing System Assurance level a.

Guidelines The emergency air supply should meet single failure criteria and be Seismic Category I.

For self-contained apparatus, there should be one extra unit for every three required.

b.

Existing FSV Conditions The Breathable Air System has two compressor / purifier trains, and it was designed and installed to Class I requirements.

There are six Scott Air Pacs installed at the Control Room where, for five men, there should be seven. There are other units in the plant, so this is not considered a deficiency.

5.

Emergency Procedures a.

Guidelines Emergency procedures to be initiated in the event of a hazardous chemical release should be written. Also, the Control Room leakage characteristics should be periodically verified.

b.

Existing FSV Conditions FSV procedure G-5 covers Personnel Emergency Responses to various accidents, including chemical spills.

This procedure essentially designates an emergency coordinator who will provide direction in the event of a hazardous chemical release.

There is no periodic control room leakage test program.

However, the amount of leakage is not considered critical to habitability because of the breathable air system and because of the charcoal filters on the ventilation

system, l

l

ATTACHMENT 4 REFERENCE ITEM III.D.3.4 b.

Existing FSV Condition i

Liquid chlorine is stored in one-ton cont:ainers approximately 1

130 meters from the control room or its air intakes. All air i

inlet and outlet dampers can be closed with switch HS75184.

Per the Union Pacific Traffic Agent, the RR track that carries liquid chlorine tankers is about 3 miles from FSV; the RR track 3/4 miles West of FSV carries mosty coal and miscellaneous freight but no chlorine.

3.

Automatic Isolation & Ventilation System Desian a.

Guidelines The control room should be protected by quick response chlorine detectors located in the fresh air inlets that will automatically close the ventilation dampers. Also, the normal fresh air makeup rate should be less than 0.3 air change per hour and the fresh air inlet should be at least 15 meters above grade.

Finally, the room should be of low leakage constr an equivalent exchange rate of less than 0.06 hr-yction, with

, and low leakage dampers should be located upstream of recirculatiun fans or at other negative pressure locations, b.

Existing FSV Conditions There are no chlorine detectors in the Control Room Ventilation System fresh air inlets; however, chlorine and other toxic materials at FSV have a strong odor that can be detected belare they build up to toxic concentrations. Chlorine is toxic at about 15 ppm, and can be smelled before 5 ppm. With the ventilation system bringing 11,400 cfm of makeup into a 40,000 cubic foot control room, it would take over three minutes to replace all the clean air with chlorinated air.

t l

Since the dampers can be isolated in 5 seconds and the respirators l

can be donned in less than 2 minutes, it is concluded that l

there would be adequate time to manually isolate the ventilation system, don respirators, and switch the ventilation system to recirculate air through charcoal filters, so that plant control would not suffer. With the control room vent inlet located 75' above grade (22 meters), it is hard to envision an accident that would introduce highly toxic chlorine concentrations into the Control Room.

Also, tge fresh air makeup rate with guts'ae dampers closed is

.F! hr, the leakage rate is.09 hr, and the dampers are bubtile tight with an 8" water differential.

These flow rates are slightly greater than recommended but meet the intent of the recommendations.

@D TRQy Room HVAC SYSTEM - Moct rito DEsisu

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X C 153 A

~

4 80 Von T VT5451 c-75!6 ro ryu l

452.

^

SID 5

500 gm.

.